Enhanced neutronics systems

ABSTRACT

Illustrative embodiments provide for the operation and simulation of the operation of fission reactors, including the movement of materials within reactors. Illustrative embodiments and aspects include, without limitation, nuclear fission reactors and reactor modules, including modular nuclear fission reactors and reactor modules, nuclear fission deflagration wave reactors and reactor modules, modular nuclear fission deflagration wave reactors and modules, methods of operating nuclear reactors and modules including the aforementioned, methods of simulating operating nuclear reactors and modules including the aforementioned, and the like.

If an Application Data Sheet (ADS) has been filed on the filing date ofthis application, it is incorporated by reference herein. Anyapplications claimed on the ADS for priority under 35 U.S.C. §§119, 120,121, or 365(c), and any and all parent, grandparent, great-grandparent,etc. applications of such applications, are also incorporated byreference, including any priority claims made in those applications andany material incorporated by reference, to the extent such subjectmatter is not inconsistent herewith.

CROSS-REFERENCE TO RELATED APPLICATIONS

The present application is related to and/or claims the benefit of theearliest available effective filing date(s) from the following listedapplication(s) (the “Priority Applications”), if any, listed below(e.g., claims earliest available priority dates for other thanprovisional patent applications or claims benefits under 35 USC §119(e)for provisional patent applications, for any and all parent,grandparent, great-grandparent, etc. applications of the PriorityApplication(s)). In addition, the present application is related to the“Related Applications,” if any, listed below.

PRIORITY APPLICATIONS

For purposes of the USPTO extra-statutory requirements, the presentapplication claims benefit of priority of U.S. Provisional PatentApplication No. 61/629,430, entitled ENHANCED NEUTRONICS SYSTEMS, namingJesse R. Cheatham III, Robert C. Petroski, Nicholas W. Touran, CharlesWhitmer as inventors, filed 18, Nov., 2011, which was filed within thetwelve months preceding the filing date of the present application or isan application of which a currently co-pending application is entitledto the benefit of the filing date.

RELATED APPLICATIONS

None

The United States Patent Office (USPTO) has published a notice to theeffect that the USPTO's computer programs require that patent applicantsreference both a serial number and indicate whether an application is acontinuation, continuation-in-part, or divisional of a parentapplication. Stephen G. Kunin, Benefit of Prior-Filed Application, USPTOOfficial Gazette Mar. 18, 2003. The USPTO further has provided forms forthe Application Data Sheet which allow automatic loading ofbibliographic data but which require identification of each applicationas a continuation, continuation-in-part, or divisional of a parentapplication. The present Applicant Entity (hereinafter “Applicant”) hasprovided above a specific reference to the application(s) from whichpriority is being claimed as recited by statute. Applicant understandsthat the statute is unambiguous in its specific reference language anddoes not require either a serial number or any characterization, such as“continuation” or “continuation-in-part,” for claiming priority to U.S.patent applications. Notwithstanding the foregoing, Applicantunderstands that the USPTO's computer programs have certain data entryrequirements, and hence Applicant has provided designation(s) of arelationship between the present application and its parentapplication(s) as set forth above and in any ADS filed in thisapplication, but expressly points out that such designation(s) are notto be construed in any way as any type of commentary and/or admission asto whether or not the present application contains any new matter inaddition to the matter of its parent application(s).

If the listings of applications provided above are inconsistent with thelistings provided via an ADS, it is the intent of the Applicant to claimpriority to each application that appears in the Priority Applicationssection of the ADS and to each application that appears in the PriorityApplications section of this application.

All subject matter of the Priority Applications and the RelatedApplications and of any and all parent, grandparent, great-grandparent,etc. applications of the Priority Applications and the RelatedApplications, including any priority claims, is incorporated herein byreference to the extent such subject matter is not inconsistentherewith.

BACKGROUND

The present application relates to nuclear fission reactors, andsystems, applications, and apparatuses related thereto.

SUMMARY

Illustrative embodiments provide for operation of nuclear fissionreactors and interfaces therewith that include simulation. Illustrativeembodiments and aspects include, without limitation, a nuclear reactormodeling interface and modeling system configured to simulate operationof a variety of nuclear fission reactors and reactor modules, includingmodular nuclear fission reactors and reactor modules, nuclear fissiondeflagration wave reactors and reactor modules, modular nuclear fissiondeflagration wave reactors and modules, methods of operating nuclearreactors and modules including the aforementioned, methods of simulatingoperating nuclear reactors and modules including the aforementioned, andthe like.

The foregoing summary is illustrative only and is not intended to be inany way limiting. In addition to the illustrative aspects, embodiments,and features described above, further aspects, embodiments, and featureswill become apparent by reference to the drawings and the followingdetailed description.

BRIEF DESCRIPTION OF THE FIGURES

The accompanying drawings, which are incorporated herein and form partof the specification, illustrate the present subject matter and,together with the description, further serve to explain the principlesof the claimed subject matter and to enable a person skilled in thepertinent art to make and use the claimed subject matter.

FIG. 1A schematically illustrates an exemplary nuclear fission reactor;

FIG. 1B is a perspective view in schematic form of an illustrativemodular nuclear fission deflagration wave reactor;

FIG. 1C schematically illustrates exemplary fluid cooling;

FIGS. 2A and 2B schematically illustrate exemplary nuclear fission fuelassemblies;

FIG. 3 schematically illustrates exemplary non-contiguous nuclearfission fuel material;

FIG. 4 schematically illustrates an exemplary modular nuclear fissionfuel core;

FIG. 5A and FIG. 5B schematically illustrate exemplary neutron affectingstructures;

FIGS. 6A and 6B schematically illustrate exemplary nuclear irradiationand movement of material;

FIGS. 7A through 7C schematically illustrate exemplary temperaturecontrol of nuclear reactivity;

FIGS. 8A through 8C schematically illustrate exemplary cells and groupsof cells;

FIG. 9 illustrates an exemplary fission yield curve;

FIG. 10 schematically illustrates an exemplary reactor control system;and

FIGS. 11-22 are flowcharts of illustrative methods associated forsimulating and/or controlling nuclear reactors.

FIG. 23 schematically illustrates a nuclear reactor modeling system.

FIG. 24A schematically illustrates a class structure of an exemplarymodeling interface.

FIG. 24B schematically illustrates an exemplary assembly structure.

FIG. 24C schematically illustrates an exemplary block structure.

FIG. 25 illustrates an exemplary input modeling data file.

FIG. 26 illustrates an exemplary input graphical user interface.

FIG. 27 illustrates an exemplary output graphical user interface.

FIGS. 28 and 29 illustrate exemplary methods.

The disclosed embodiments will now be described with reference to theaccompanying drawings. In the drawings, like reference numbers mayindicate identical or similar elements. Additionally, the left-mostdigit(s) of a reference number may identify the drawing in which thereference number first appears.

DETAILED DESCRIPTION Introduction

In the following detailed description, reference is made to theaccompanying drawings, which form a part hereof. In the drawings,generally similar symbols identify similar components, unless contextdictates otherwise. The illustrative embodiments described in thedetailed description, drawings, and claims are not meant to be limiting.Other embodiments may be utilized, and other changes may be made,without departing from the spirit or scope of the subject matterpresented here.

It is to be appreciated that the Detailed Description section, and notthe Summary and Abstract sections, is intended to be used to interpretthe claims. The Summary and Abstract sections may set forth one or morebut not all exemplary embodiments and, thus, are not intended to limitthe claimed subject matter and the appended claims in any way.

While specific configurations and arrangements are discussed, it shouldbe understood that this is done for illustrative purposes only. A personskilled in the pertinent art will recognize that other configurationsand arrangements can be used without departing from the spirit and scopeof the claimed subject matter. It will be apparent to a person skilledin the pertinent art that the claimed subject matter can also be used ina variety of other applications. The scope of the claimed subject matteris not limited to the disclosed embodiments. The claimed subject matteris defined by the claims appended hereto.

References to “one embodiment,” “an embodiment,” “this embodiment,” “anexample embodiment,” etc., indicate that the embodiment described mayinclude a particular feature, structure, or characteristic, but everyembodiment might not necessarily include the particular feature,structure or characteristic. Moreover, such phrases are not necessarilyreferring to the same embodiment. Further, when a particular feature,structure, or characteristic is described in connection with anembodiment, it is understood that it is within the knowledge of oneskilled in the art to effect such a feature, structure, orcharacteristic in connection with other embodiments whether or notexplicitly described.

In some instances, one or more components may be referred to herein as“configured to,” “configurable to,” “operable/operative to,”“adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Thoseskilled in the art will recognize that such terms (e.g. “configured to”)can generally encompass active-state components and/or inactive-statecomponents and/or standby-state components, unless context requiresotherwise.

Those having skill in the art will recognize that the state of the arthas progressed to the point where there is little distinction leftbetween hardware, software, and/or firmware implementations of aspectsof systems; the use of hardware, software, and/or firmware is generally(but not always, in that in certain contexts the choice between hardwareand software can become significant) a design choice representing costvs. efficiency tradeoffs. Those having skill in the art will appreciatethat there are various vehicles by which processes and/or systems and/orother technologies described herein can be effected (e.g., hardware,software, and/or firmware), and that the preferred vehicle will varywith the context in which the processes and/or systems and/or othertechnologies are deployed. For example, if an implementer determinesthat speed and accuracy are paramount, the implementer may opt for amainly hardware and/or firmware vehicle; alternatively, if flexibilityis paramount, the implementer may opt for a mainly softwareimplementation; or, yet again alternatively, the implementer may opt forsome combination of hardware, software, and/or firmware. Hence, thereare several possible vehicles by which the processes and/or devicesand/or other technologies described herein may be effected, none ofwhich is inherently superior to the other in that any vehicle to beutilized is a choice dependent upon the context in which the vehiclewill be deployed and the specific concerns (e.g., speed, flexibility, orpredictability) of the implementer, any of which may vary. Those skilledin the art will recognize that optical aspects of implementations willtypically employ optically-oriented hardware, software, and or firmware.

The foregoing detailed description has set forth various embodiments ofdevices and/or processes via the use of block diagrams, flowcharts,and/or examples. Insofar as such block diagrams, flowcharts, and/orexamples contain one or more functions and/or operations, it will beunderstood by those within the art that each function and/or operationwithin such block diagrams, flowcharts, or examples can be implemented,individually and/or collectively, by a wide range of hardware, software,firmware, or virtually any combination thereof. In an embodiment,several portions of the subject matter described herein may beimplemented via Application Specific Integrated Circuits (ASICs), FieldProgrammable Gate Arrays (FPGAs), digital signal processors (DSPs), orother integrated formats. However, those skilled in the art willrecognize that some aspects of the embodiments disclosed herein, inwhole or in part, can be equivalently implemented in integratedcircuits, as one or more computer programs running on one or morecomputers (e.g., as one or more programs running on one or more computersystems), as one or more programs running on one or more processors(e.g., as one or more programs running on one or more microprocessors),as firmware, or as virtually any combination thereof, and that designingthe circuitry and/or writing the code for the software and or firmwarewould be well within the skill of one of skill in the art in light ofthis disclosure. In addition, those skilled in the art will appreciatethat the mechanisms of the subject matter described herein are capableof being distributed as a program product in a variety of forms, andthat an illustrative embodiment of the subject matter described hereinapplies regardless of the particular type of signal bearing medium usedto actually carry out the distribution. Examples of a signal bearingmedium include, but are not limited to, the following: a recordable typemedium such as a floppy disk, a hard disk drive, a Compact Disc (CD), aDigital Video Disk (DVD), a digital tape, a computer memory, etc.; and atransmission type medium such as a digital and/or an analogcommunication medium (e.g., a fiber optic cable, a waveguide, a wiredcommunications link, a wireless communication link (e.g., transmitter,receiver, transmission logic, reception logic, etc.), etc.).

By way of overview, illustrative embodiments provide nuclear fissionreactors, and apparatuses and methods for their operation andsimulation. Illustrative embodiments and aspects include, withoutlimitation, nuclear fission reactors and reactor modules, includingmodular nuclear fission reactors and reactor modules, nuclear fissiondeflagration wave reactors and reactor modules, modular nuclear fissiondeflagration wave reactors and modules, methods of operating nuclearreactors and modules including the aforementioned, methods of simulatingoperating nuclear reactors and modules including the aforementioned, andthe like.

Still by way of overview and referring to FIG. 1A, an illustrativenuclear fission reactor 10 will be discussed by way of illustration andnot limitation. Nuclear fission reactor 10 may be, but is not limitedto, a fission deflagration wave reactor. The reactor 10 suitablyincludes a nuclear reactor core 100 disposed within a reactor vessel 12and a reactor coolant system having one or more reactor coolant loops14.

A reactor may be a modular design including one or more nuclear reactormodules-see, e.g., an exemplary modular reactor 50 illustrated in FIG.1B. Each reactor module 12 may be operatively coupled in fluidcommunication to at least one heat sink 58 via a reactor coolant system56. Thus, each of the nuclear reactor modules may be considered acomplete, stand-alone nuclear reactor system by itself. A nuclearreactor module may be neutronically coupled to at least one otheradjacent reactor module. Thus, adjacent nuclear reactor modules can beneutronically integrated yet physically separate from each other.

In order to provide an understanding of the control and simulation ofreactors such as reactor 10 and reactor 50, illustrative corenucleonics, given by way of non-limiting examples, will be set forthfirst. While many reactor embodiments are contemplated, several of thesenon-limiting examples are illustrated in U.S. patent application Ser.No. 12/069,907 entitled MODULAR NUCLEAR FISSION REACTOR, naming AHLFELD,CHARLES E., GILLELAND, JOHN ROGERS, HYDE, RODERICK A., ISHIKAWA, MURIELY., MCALEES, DAVID G., MYHRVOLD, NATHAN P., WHITMER, CHARLES, and WOOD,LOWELL L. as inventors, filed 12 Feb. 2008, U.S. patent application Ser.No. 11/605,943, entitled AUTOMATED NUCLEAR POWER REACTOR FOR LONG-TERMOPERATION, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P.MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, U.S.patent application Ser. No. 11/605,848, entitled METHOD AND SYSTEM FORPROVIDING FUEL IN A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y.ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors,filed 28 Nov. 2006, and U.S. patent application Ser. No. 11/605,933,entitled CONTROLLABLE LONG TERM OPERATION OF A NUCLEAR REACTOR, namingRODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L.WOOD, JR. as inventors, filed 28 Nov. 2006, the entire contents of whichare hereby incorporated by reference in their entireties. Then, detailswill be set forth regarding several illustrative embodiments and aspectsof reactors.

Considerations

Before discussing details of the reactors such as reactor 10 and reactor50, some considerations behind reactor embodiments will be given by wayof overview but are not to be interpreted as limitations. Some reactorembodiments address many of the considerations discussed below. On theother hand, some other reactor embodiments may address one or a selectfew of these considerations, and need not accommodate all of theconsiderations discussed below.

Certain of the nuclear fission fuels envisioned for use in reactorembodiments are typically widely available, such as without limitationuranium (natural, depleted, or enriched), thorium, plutonium, or evenpreviously-burned nuclear fission fuel assemblies. Other, less widelyavailable nuclear fission fuels, such as without limitation otheractinide elements or isotopes thereof may be used in embodiments of thereactor. While some reactor embodiments contemplate long-term operationat full power, or some portion thereof, on the order of around ⅓ centuryto around ½ century or longer, an aspect of some reactor embodimentsdoes not contemplate nuclear refueling. Other reactor embodimentscontemplate nuclear refueling, however. In some cases, embodiments maycontemplate burial in-place at end-of-life. Nuclear refueling may occurduring shutdown periods and/or operation at power. It is alsocontemplated that nuclear fission fuel reprocessing may be avoided insome cases, thereby mitigating possibilities for diversion to militaryuses and other issues.

Some reactor embodiments may be sited underground, thereby addressinglarge, abrupt releases and small, steady-state releases of radioactivityinto the biosphere. Some embodiments may entail minimizing operatorcontrols, thereby automating those embodiments as much as practicable.In some embodiments, a life-cycle-oriented design is contemplated,wherein those embodiments can operate from startup to shutdown atend-of-life. In some life-cycle oriented designs, the embodiments mayoperate in a substantially fully-automatic manner. Some embodiments lendthemselves to modularized construction. Finally, some embodiments may bedesigned according to high power density or to selected power densitiescorresponding to a variety of design considerations, such as burn-upcriteria, power demand, neutronic flux considerations, and otherparameters.

During operation, the materials (e.g., elements and isotopes ofelements) in a reactor, especially a reactor core region, change overtime. For example, fuel atoms fission into fission products. Atoms offuel, structural materials, neutron absorbing materials (fission productpoisons or neutron absorbing materials intentionally inserted into thereactor), and so forth may absorb neutrons and become other isotopes orelements. These changes may be accounted for by design and reactorcontrol in both the short term and the long term. An ability to movematerials throughout the core may increase a reactor's effectivelifetime.

Some features of various reactor embodiments result from some of theabove considerations. For example, simultaneously accommodating desiresto achieve ⅓/½ century (or longer) of operations at full power withoutshutdown for nuclear refueling and to avoid nuclear fission fuelreprocessing may entail use of a fast neutron spectrum. As anotherexample, in some embodiments a negative temperature coefficient ofreactivity (αT) is engineered-in to the reactor, such as via negativefeedback on local reactivity implemented with strong absorbers ofneutrons or other approaches to reactivity control. In the alternativeor in addition, some embodiments are configured to control the fissionprocess in whole or in part by achieving a spectral shift in a neutronflux using spectral control methods such as displacing and/or insertinga neutron moderator for some time period. As a further example, in somemodular deflagration wave embodiments, a distributed thermostat enablesa propagating nuclear fission deflagration wave mode of nuclear fissionfuel burn. This mode simultaneously permits a high average burn-up ofnon-enriched actinide fuels, such as natural uranium or thorium, and useof a comparatively small “nuclear fission igniter” region of moderateisotopic enrichment of nuclear fissionable materials in the core's fuelcharge. As another example, in some embodiments, multiple redundancy isprovided in primary and secondary core cooling.

Exemplary Embodiments of Nuclear Fission Reactors

Now that some of the considerations behind some of the reactorembodiments have been set forth, further details regarding an exemplaryembodiment of nuclear fission reactors will be explained. Thisinformation is provided to enhance the understanding of theconsiderations taken into account when modeling and simulaitng nuclearreactor performance. It is emphasized that the following description ofexemplary nuclear reactor embodiments is given by way of non-limitingexamples only and not by way of limitation. As mentioned above, severalembodiments of nuclear reactors and their simulation are contemplated,as well as further aspects of reactor 10. After details regarding anexemplary embodiment of reactor 10 are discussed, other embodiments andaspects will also be discussed.

Still referring to FIG. 1A, an exemplary embodiment of reactor 10includes a reactor core assembly 100 that is disposed within a reactorpressure vessel 12. Several embodiments and aspects of reactor coreassembly 100 are contemplated that will be discussed later. Some of thefeatures that will be discussed later in detail include nuclear fissionfuel materials and their respective nucleonics, fuel assemblies, fuelgeometries, and the operation and simulation of reactor core assembly100 in a complete reactor system.

The reactor pressure vessel 12 suitably is any acceptable pressurevessel known in the art and may be made from any appropriate form ofmaterials acceptable for use in reactor pressure vessels, such aswithout limitation stainless steel. Within the reactor pressure vessel12, a neutron reflector (not shown) and a radiation shield (not shown)surround reactor core assembly 100. In some embodiments, the reactorpressure vessel 12 is sited underground. In such cases, the reactorpressure vessel 12 can also function as a burial cask for reactor coreassembly 100. In these embodiments, the reactor pressure vessel 12suitably is surrounded by a region (not shown) of isolation material,such as dry sand, for long-term environmental isolation. The region (notshown) of isolation material may have a size of around 100 meters indiameter or so. However, in other embodiments, the reactor pressurevessel 12 is sited on or toward the Earth's surface.

Reactor coolant loops 14 transfer heat from nuclear fission in reactorcore assembly 100 to application heat exchangers 16. The reactor coolantmay be selected as desired for a particular application. In someembodiments, the reactor coolant suitably is helium (He) gas. In otherembodiments, the reactor coolant suitably may be other pressurized inertgases, such as neon, argon, krypton, xenon, or other fluids such aswater or gaseous or superfluidic carbon dioxide, or liquid metals, suchas sodium or lead, or metal alloys, such as Pb—Bi, or organic coolants,such as polyphenyls, or fluorocarbons. The reactor coolant loopssuitably may be made from tantalum (Ta), tungsten (W), aluminum (Al),steel or other ferrous or non-iron groups alloys or titanium orzirconium-based alloys, or from other metals and alloys, or from otherstructural materials or composites, as desired.

In some embodiments, the application heat exchangers 16 may be steamgenerators that generate steam that is provided as a prime mover forrotating machinery, such as electrical turbine-generators 18 within anelectrical generating station 20. In such a case, reactor core assembly100 suitably operates at a high operating pressure and temperature, suchas above 1,000K or so and the steam generated in the steam generator maybe superheated steam. In other embodiments, the application heatexchanger 16 may be any steam generator that generates steam at lowerpressures and temperatures (that is, need not be superheated steam) andreactor core assembly 100 operates at temperatures less than around550K. In these cases, the application heat exchangers 16 may provideprocess heat for applications such as desalination plants for seawateror for processing biomass by distillation into ethanol, or the like.

Optional reactor coolant pumps 22 circulate reactor coolant throughreactor core assembly 100 and the application heat exchangers 16. Notethat although the illustrative embodiment shows pumps andgravitationally driven circulation, other approaches may not utilizepumps, or circulatory structures or be otherwise similarly geometricallylimited. The reactor coolant pumps 22 suitably may be provided whenreactor core assembly 100 is sited approximately vertically coplanarwith the application heat exchangers 16, such that thermal driving headis not generated. The reactor coolant pumps 22 may also be provided whenreactor core assembly 100 is sited underground. However, when reactorcore assembly 100 is sited underground or in any fashion so reactor coreassembly 100 is vertically spaced below the application heat exchangers16, thermal driving head may be developed between the reactor coolantexiting the reactor pressure vessel 12 and the reactor coolant exitingthe application heat exchangers 16 at a lower temperature than thereactor coolant exiting the reactor pressure vessel 12. When sufficientthermal driving head exists, reactor coolant pumps 22 need not beprovided to provide sufficient circulation of reactor coolant throughreactor core assembly 100 to remove heat from fission during operationat power.

In some embodiments more than one reactor coolant loop 14 may beprovided, thereby providing redundancy in the event of a casualty, suchas a loss of coolant accident, a loss of flow accident, aprimary-to-secondary leak or the like, to any one of the other reactorcoolant loops 14. Each reactor coolant loop 14 may be rated forfull-power operation, though some applications may remove thisconstraint.

In some embodiments, closures 24, such as reactor coolant shutoffvalves, are provided in lines of the reactor coolant system 14. In eachreactor coolant loop 14, a closure 24 may be provided in an outlet linefrom the reactor pressure vessel 12 and in a return line to the reactorpressure vessel 12 from an outlet of the application heat exchanger 16.Closures 24 may be fast-acting closures that shut quickly underemergency conditions, such as detection of significant fission-productentrainment in reactor coolant. Closures 24 may be provided in additionto a redundant system of automatically-actuated valves (not shown).

One or more heat-dump heat exchangers 26 are provided for removal ofafter-life heat (decay heat). Heat-dump heat exchanger 26 includes aprimary loop that is configured to circulate decay heat removal coolantthrough reactor core assembly 100. Heat-dump heat exchanger 26 includesa secondary loop that is coupled to an engineered heat-dump heat pipenetwork (not shown). In some situations, for example, for redundancypurposes, more than one heat-dump heat exchanger 26 may be provided.Heat-dump heat exchanger 26 may be sited at a vertical distance abovereactor core assembly 100 so sufficient thermal driving head is providedto enable natural flow of decay heat removal coolant without need fordecay heat removal coolant pumps. However, in some embodiments decayheat removal pumps (not shown) may be provided. Reactor coolant pumpsmay be used for decay heat removal, where appropriate.

Now that an overview of an exemplary embodiment of the reactor 10 hasbeen given, other embodiments and aspects will be discussed. First,embodiments and aspects of reactor core assembly 100 will be discussed.An overview of reactor core assembly 100 and its nucleonics will be setforth first, followed by descriptions of exemplary embodiments and otheraspects of reactor core assembly 100. Again, this information enhancesthe understanding and considerations taken into account when modeling orsimulating nuclear reactor performance.

Given by way of overview and in general terms, structural components ofreactor core assembly 100 may be made of tantalum (Ta), tungsten (W),rhenium (Re), various alloys including but not limited to steels such asmartensitic stainless steels (e.g., HT9), austenitic stainless steels(e.g., Type 316), or carbon composite, ceramics, or the like. Thesematerials are suitable because of the high temperatures at which reactorcore assembly 100 operates, and because of their creep resistance overthe envisioned lifetime of full power operation, mechanical workability,and corrosion resistance. Structural components can be made from singlematerials, or from combinations of materials (e.g., coatings, alloys,multilayers, composites, and the like). In some embodiments, reactorcore assembly 100 operates at sufficiently lower temperatures so thatother materials, such as aluminum (Al), steel, titanium (Ti) or the likecan be used, alone or in combinations, for structural components.

In deflagration wave embodiments, reactor core assembly 100 may includea small nuclear fission igniter and a larger nuclear fissiondeflagration burn-wave-propagating region. The nuclear fissiondeflagration burn-wave-propagating region suitably contains thorium oruranium fuel, and functions on the general principle of fast neutronspectrum fission breeding. In some deflagration wave embodiments,uniform temperature throughout reactor core assembly 100 is maintainedby thermostating modules which regulate local neutron flux and therebycontrol local power production. Some example deflagration waveembodiments are further discussed in the aforementioned U.S. patentapplication Ser. No. 11/605,933, entitled CONTROLLABLE LONG TERMOPERATION OF A NUCLEAR REACTOR (“the '933 application”), which is hereinincorporated by reference in its entirety.

Nuclear reactors may be modular. Referring now to FIG. 1B, anillustrative modular reactor 50 is shown. It is emphasized that thefollowing description of an exemplary embodiment of reactor 50 is givenby way of non-limiting example only and not by way of limitation. Asmentioned above, several embodiments of reactors such as reactors 10 and50, are contemplated, as well as further aspects of reactors. Featuresillustrated in reactors 10 and 50 may be implemented separately or inany suitable combination. After details regarding an exemplaryembodiment of reactor 50 are discussed, other embodiments and aspectswill also be discussed.

Modular reactor 50 is shown by way of illustration and does not limitmodular reactors to a toroidal arrangement or any other arrangement ofreactor modules 52. It will be understood that no limitation to such ageometric arrangement or to any geometric arrangement of any typewhatsoever is intended. To that end, additional arrangements of reactormodules 52 will be discussed further below. In the interest of brevity,the description of additional arrangements of reactor modules 52 islimited to those illustrated herein. However, it will be appreciatedthat reactor modules 52 may be arranged in any manner whatsoever asdesired and may accommodates neutronic coupling of adjacent nuclearfission deflagration wave reactor modules 52.

As discussed above, the exemplary modular reactor 50 suitably includesreactor modules 52. Each reactor module 52 may suitably include areactor core 54 and a reactor coolant system 56. Each nuclear fissiondeflagration wave reactor module 52 may be operatively coupled in fluidcommunication to at least one heat sink 58 via one or more associatedreactor coolant systems 56. That is, each reactor modules 52 suitablymay be considered a complete, stand-alone nuclear reactor by itself. Areactor module 52 may be neutronically coupled to at least one adjacentreactor module 52. While many embodiments of the modular reactor 50 arecontemplated, a common feature among many contemplated embodiments ofmodular reactor 50 is neutronic coupling of adjacent reactor modules 52via origination of a nuclear fission deflagration wave, or “bumfront” asfurther discussed in the aforementioned U.S. patent application Ser. No.12/069,907 entitled MODULAR NUCLEAR FISSION REACTOR (the '907application”), which is herein incorporated by reference in itsentirety.

Referring now to FIG. 1C, heat energy can be extracted from a nuclearfission reactor core according to another embodiment. In a nuclearfission reactor 110, nuclear fission occurs in a heat generating region120 (e.g., throughout the fuel-bearing core or propagated in a burningwavefront, for example). Heat absorbing material 160, such as acondensed phase density fluid (e.g., water, liquid metals, terphenyls,polyphenyls, fluorocarbons, FLIBE (2LiF—BeF₂) and the like) flowsthrough the region 120 as indicated by an arrow 150, and heat istransferred from the heat generating region 120 to heat absorbingmaterial 160. In some embodiments, e.g., fast fission spectrum nuclearreactors, heat absorbing material 160 is chosen to be a nuclear inertmaterial (such as He4) so as to minimally perturb the neutron spectrum.In other embodiments of nuclear fission reactor 110, the neutron contentis sufficiently robust, so that a non-nuclear-inert heat absorbingmaterial 160 may be acceptably utilized. Heat absorbing material 160flows (e.g., by natural convection or by forced movement) to a heatextraction region 130 that is substantially out of thermal contact withheat generating region 120. Heat energy 140 is extracted from heatabsorbing material 160 at heat extraction region 130. Heat absorbingmaterial 110 can reside in either a liquid state, a multiphase state, ora substantially gaseous state upon extraction of the heat energy 140 inthe heat extraction region 130.

Exemplary Movements of Nuclear Reactor Materials

Fuel materials include not only fuel materials, but also structuralmaterials (e.g., cladding). Referring now to FIG. 2A, a reactor 200,which may include any type of fission reactor including those describedelsewhere herein, may include nuclear fission fuel assemblies 210disposed therein. The following discussion includes details of exemplarynuclear fission fuel assemblies 210 that may be used in reactor 200.Referring now to FIG. 2B and given by way of non-limiting example, in anembodiment the nuclear fission fuel assembly 210 suitably includes anuclear fission fuel assembly 220. In an embodiment, nuclear fissionfuel assembly 220 has been “previously burnt.” The term “previouslyburnt” means that at least some components of the nuclear fission fuelassembly have undergone neutron-mediated nuclear fission and that theisotopic composition of the nuclear fission fuel has been modified. Thatis, the nuclear fission fuel assembly has been put in a neutron spectrumor flux (either fast or slow), at least some components have undergoneneutron-mediated nuclear fission and, as result, the isotopiccomposition of the nuclear fission fuel has been changed. Thus, apreviously burnt nuclear fission fuel assembly 220 may have beenpreviously burnt in any reactor including reactor 200, such as withoutlimitation a light water reactor. Previously burnt fission fuel (e.g.,in a previously burnt nuclear fission fuel assembly 220) may bechemically untreated subsequent to its previous burning.

It is intended that nuclear fission fuel assembly 220 can includewithout limitation any type of nuclear fissionable material whatsoeverappropriate for undergoing fission in a nuclear fission reactor, such asactinide or transuranic elements like natural thorium, natural uranium,enriched uranium, or the like. Nuclear fission fuel assembly 220 is cladwith cladding 224. If nuclear fission fuel assembly has been previouslyburnt, the cladding 224 may be the “original” cladding in which thenuclear fission fuel assembly 220 was clad before it was burnt. In someother embodiments, a previously burnt nuclear fission fuel assembly 220may not be clad with “original” cladding 224. For example, a previouslyburnt nuclear fission fuel assembly 220 may be retained in its originalcladding 224, and a new cladding (not shown) may be disposed around anexterior of cladding 224. In some embodiments, the new cladding is madeup of cladding sections (not shown) that are configured to helpaccommodate swelling into the void spaces. In other embodiments, the newcladding may be provided as a barrier, such as a tube, provided betweenan exterior of the cladding 224 and reactor coolant (not shown).

Referring now to FIG. 3, an exemplary nuclear fission fuel structure 300includes non-contiguous segments 320 of nuclear fission fuel material.Non-contiguous segments 320 may be in “neutronic” contact without beingin physical contact. Nuclear fission fuel structure 300 may also includean optional nuclear fission igniter 310. As described in theaforementioned '933 application, nuclear fission igniter 310 may be usedin deflagration propagating wave-type nuclear reactors.

Referring now to FIG. 4, a modular nuclear fission fuel core 400 mayinclude an optional neutron reflector/radiation shield 410 and modularassemblies 420. Modular assemblies 420 may be modular fuel assemblieshaving some fuel material content. Modular assemblies may also bemodular neutron absorbing assemblies (having some neutron absorbingmaterial content), modular structural assemblies (serving a primarilystructural purpose), modular payload assemblies (designed to carry apayload of, for example, a material to be subjected to a neutron flux),modular blank assemblies (serving as a mere placeholder, for example, toreduce the nucleonic, flow, structural, and thermal perturbationsinduced by a void or void filled with coolant and/or moderator), or anycombination of the above.

Modular assemblies 420 are placed as desired within the assemblyreceptacles 430. Modular nuclear fission fuel core 400 may be operatedin any number of ways. For example, all of the assembly receptacles 430in the modular nuclear fission fuel core 400 may be fully populated withmodular fuel assemblies 420 prior to initial operation. For example, indeflagration propagating wave-type nuclear reactor embodiments, prior toinitial operation means prior to origination and propagation of anuclear fission deflagration propagating wave burnfront within andthrough the modular fuel assemblies 420. In other reactor embodiments,prior to initial operation means prior to initial criticality or priorto a modular nuclear fission fuel core being exposed to a neutron flux.

As another example, modular assemblies 420 may be removed from theirrespective assembly receptacles 430 and replaced with other modularassemblies 440 (of the same or different type), as desired; thisemplacement is indicated by the arrow 444. For example, “burnt” fuelassemblies may be replaced with “unburnt” fuel assemblies, neutronabsorbing assemblies may be replaced with fuel assemblies, and so forth.The other modular nuclear assemblies 440 may be unused or may havepreviously been used. For example, in deflagration propagating wave-typenuclear reactor embodiments, modular fission fuel assemblies 420 may beremoved and replaced with other modular nuclear fission fuel assemblies440 after a nuclear fission deflagration wave burnfront has completelypropagated through modular nuclear fission fuel assemblies 420. In otherembodiments, modular assemblies 420 may be removed and replaced withother modular assemblies 440 for any reason (e.g., testing orexperimental uses, redistribution of fuel or neutron absorbingmaterials, etc.). Such replacement strategies may be used to extendoperation of modular nuclear fission fuel core 400 as desired.

As another example, the modular nuclear fission fuel core 400 need notbe fully populated with modular assemblies 420 prior to initialoperation. For example, less than all of the assembly receptacles 430can be populated with modular assemblies 420. In such a case, the numberof modular fuel assemblies that are placed within the modular nuclearfission fuel core 400 can be determined based upon many reasons, such asa number of modular fuel assemblies that are available, power demand(e.g., electrical loading in watts), that will be ultimately be placedupon the modular nuclear fission fuel core 400, etc. Thus, continued orextended operation of the modular nuclear fission fuel core 400 can beenabled without initially fueling the entire modular nuclear fissionfuel core 400 with modular fuel assemblies.

It will be appreciated that the concept of modularity can be extended.For example, in other embodiments, a modular nuclear fission reactor canbe populated with any number of nuclear fission reactor cores in thesame manner that the modular nuclear fission fuel core 400 can bepopulated with any number of modular assemblies 420. To that end, themodular nuclear fission reactor can be analogized to the modular nuclearfission fuel core 400 and nuclear fission reactor cores can beanalogized to the modular nuclear fission fuel assemblies 420. Theseveral contemplated modes of operation discussed above for the modularnuclear fission fuel core 400 thus apply by analogy to a modular nuclearfission reactor.

Core materials not in a modular assembly may also be moved in a reactorcore. It is well known in the art to control reactivity (and thus coreaverage temperature in an operating reactor having a negativecoefficient of reactivity) using control rods or other devices. Inaddition, other neutron modifying structures are contemplated inembodiments. For example, referring now to FIGS. 5A and 5B, neutronmodifying structures 530 can position neutron modifying (e.g.,absorbing, reflecting, moderating, etc.) substances in a reactor 500,including a propagating burnfront nuclear fission reactor 550, for avariety of purposes. In an embodiment, neutron modifying structures 530insert neutron absorbers, such as without limitation Li-6, B-10, or Gd,into nuclear fission fuel. In another embodiment, neutron modifyingstructures 530 insert neutron moderators, such as without limitationhydrocarbons or Li-7, thereby modifying the neutron energy spectrum, andthereby changing the neutronic reactivity of nuclear fission fuel in thelocal region.

In some situations in a reactor 500 (including a propagating burnfrontnuclear fission reactor 550) an effect of the neutron moderators isassociated with detailed changes in the neutron energy spectrum (e.g.,hitting or missing cross-section resonances), while in other cases theeffects are associated with lowering the mean neutron energy of theneutron environment (e.g., downshifting from “fast” neutron energies toepithermal or thermal neutron energies). In yet other situations, aneffect of the neutron moderators is to deflect neutrons to or away fromselected locations. In some embodiments, one of the aforementionedeffects of neutron moderators is of primary importance, while in otherembodiments, multiple effects are of comparable or lesser designsignificance. In another embodiment, neutron modifying structures 530contain both neutron absorbers and neutron moderators; in onenonlimiting example, the location of neutron absorbing material relativeto that of neutron moderating material is changed to affect control(e.g., by masking or unmasking absorbers, or by spectral-shifting toincrease or decrease the absorption of absorbers), in anothernonlimiting example, control is affected by changing the amounts ofneutron absorbing material and/or neutron moderating material.

In embodiments such as propagating burnfront nuclear fission reactor550, a nuclear fission deflagration wave burnfront can be driven intoareas of nuclear fission fuel as desired, thereby enabling a variablenuclear fission fuel burn-up. In propagating burnfront nuclear fissionreactor 550, a nuclear fission deflagration wave burnfront 510 isinitiated and propagated. Neutron modifying structures 530 can direct ormove the burnfront 510 in directions indicated by arrows 520. In anembodiment, neutron modifying structures 530 insert neutron absorbersbehind burnfront 510, thereby driving down or lowering neutronicreactivity of fuel that is presently being burned by burnfront 510relative to neutronic reactivity of fuel ahead of burnfront 510, therebyspeeding up the propagation rate of the nuclear fission deflagrationwave. In another embodiment, neutron modifying structures 530 insertneutron absorbers into nuclear fission fuel ahead of burnfront 510,thereby slowing down the propagation of the nuclear fission deflagrationwave. In other embodiments, neutron modifying structures 530 insertneutron absorbers into nuclear fission fuel within or to the side of theburnfront 510, thereby changing the effective size of the burnfront 510.In another embodiment, neutron modifying structures 530 insert neutronmoderators, thereby modifying the neutron energy spectrum, and therebychanging the neutronic reactivity of nuclear fission fuel that ispresently being burned by the burnfront 510 relative to neutronicreactivity of nuclear fission fuel ahead of or behind the burnfront 510.

Thus, local neutronic reactivity in reactor 500, and burnfront 510 inpropagating burnfront nuclear fission reactor 550, can be directed asdesired according to selected local reaction rate or propagationparameters. For example, local reaction rate parameters can includefission rate, a heat generation density, cross-section dimensions ofpower density, or the like. In burnfront nuclear fission reactor 550,propagation parameters can include a propagation direction ororientation of the burnfront 510, a propagation rate of the burnfront510, power demand parameters such the heat generation density,cross-sectional dimensions of a burning region through which theburnfront 510 is to the propagated (such as an axial or lateraldimension of the burning region relative to an axis of propagation ofthe burnfront 510), or the like. For example, the propagation parametersmay be selected so as to control the spatial or temporal location of theburnfront 510, so as to avoid failed or malfunctioning control elements(e.g., neutron modifying structures or thermostats), or the like.

Neutron modifying structures 530 may be actively controlled and/orpassively controlled (e.g., programmable). Actively controlled neutronmodifying structures are actively controlled by an operator and/or anexternal control system. Passively controlled neutron modifyingstructures are responsive to conditions at one or more locations in thecore. For example, programmable temperature responsive neutron modifyingstructures (examples of which are discussed in detail in theaforementioned '933 application) introduce and remove neutron absorbingor neutron moderating material into and from the fuel-charge of areactor 500 (including embodiments such as propagating burnfront nuclearfission reactor 550). Responsive to an operating temperature profile,programmable temperature responsive neutron modifying structuresintroduce neutron absorbing or moderating material into the fuel-chargeof the nuclear fission reactor to lower operating temperature in thenuclear fission reactor or remove neutron absorbing or moderatingmaterial from the fuel-charge of the nuclear fission reactor in order toraise operating temperature of the nuclear fission reactor.

It will be appreciated that temperatures are only one example of controlparameters which can be used to determine the control settings ofpassively controlled or programmable neutron modifying structures.Nonlimiting examples of other control parameters which can be used todetermine the control settings of programmable neutron modifyingstructures include power levels, neutron levels, neutron spectrum,neutron absorption, fuel burnup levels, and the like. In one example,the neutron modifying structures are used to control fuel burnup levelsto relatively low (e.g., <50%) levels in order to achieve high-rate“breeding” of nuclear fission fuel for use in other nuclear fissionreactors, or to enhance suitability of the burnt nuclear fission fuelfor subsequent re-propagation of a nuclear fission deflagration wave ina propagating nuclear fission deflagration wave reactor. Differentcontrol parameters can be used at different times, or in differentportions of the reactor. It will be appreciated that the various neutronmodifying methods discussed previously in the context of neutronmodifying structures can also be utilized in programmable temperatureresponsive neutron modifying structures, including without limitation,the use of neutron absorbers, neutron moderators, combinations ofneutron absorbers and/or neutron moderators, variable geometry neutronmodifiers, and the like.

A material may be subjected to a neutron flux in a reactor. It should beappreciated that the neutron irradiation of material in a reactor may becontrolled by the duration and/or extent of duration and local powerlevel. In another embodiment, the neutron irradiation of material may becontrolled by control of the neutron environment (e.g., the neutronenergy spectrum for Np-237 processing) via neutron modifying structures.Referring to FIGS. 6A and 6B, for example, a material 610 inserted intoa reactor 600, as indicated generally with arrow 602, will be subject toa neutron flux dependent upon, inter alia, local power level, duration,neutron modifying structures, and/or neutron spectrum modifyingfeatures. In an embodiment where the reactor is a propagating nuclearfission deflagration wave reactor, such as reactor 650, material 610 maybe inserted into reactor 650 as indicated generally with arrow 652. Inanother embodiment, propagating nuclear fission deflagration wavereactor 650 may be operated in a “safe” sub-critical manner, relyingupon an external source of neutrons to sustain the propagatingburnfront, while using a portion of the fission-generated neutrons fornuclear processing of core materials. It should be appreciated that themovement of material 610 to a location within reactor 600 (or 650) maybe from a location external to the reactor (as shown) or from anotherlocation within the reactor (not shown).

In some embodiments, a material 610 may be present in a location withinthe reactor before nuclear fission ignition occurs within a reactor,while in other embodiments the material may be added (i.e., moved to thelocation) after nuclear fission occurs or occurs in that locale. In someembodiments, material is removed from the reactor, while in otherembodiments it remains in place. Alternately, a material having a set ofnon-irradiated properties is loaded into a reactor. The material istransported (e.g., as indicated generally by arrows 652 and 602) intophysical proximity and neutronic coupling with a region of maximizedreactivity—in the case of propagating nuclear fission deflagration wavereactor 650, as the nuclear fission deflagration wave propagatingburnfront (e.g., burnfront 670) passes through the material.

The material 610 remains in neutronic coupling for a sufficient timeinterval to convert the material 610 into a second material 606 having adesired set of modified properties. Upon the material 610 having thusbeen converted into the material 606, the material 606 may be physicallytransported out of reactor 600 (or reactor 650) as generally indicatedby arrow 604 (or 654). The removal can take place either duringoperation of reactor 600 (or 650) or after shutdown. The removal can beperformed as a continuous, sequential, or batch process. In one example,nuclearly processed material 606 may be subsequently used as nuclearfission fuel in another nuclear fission reactor, such as withoutlimitation LWRs or propagating nuclear fission deflagration wavereactors. In another nonlimiting example, nuclearly processed material606 may be subsequently used within the nuclear fission ignitor of apropagating nuclear fission deflagration wave reactor. In one approach,thermal management may be adjusted to provide thermal controlappropriate for any changes in operational parameters, as appropriatefor the revised materials or structures.

According to further embodiments, temperature-driven neutron absorptioncan be used to control a nuclear fission reactor, thereby“engineering-in” an inherently-stable negative temperature coefficientof reactivity (aT). Referring now to FIG. 7A, a nuclear reactor 700 isinstrumented with temperature detectors 710, such as without limitationthermocouples. In this embodiment, the nuclear fission reactor 700suitably can be any type of fission reactor whatsoever. To that end, thenuclear fission reactor 700 can be a thermal neutron spectrum nuclearfission reactor or a fast neutron spectrum nuclear fission reactor, asdesired for a particular application.

For example, temperature detectors detect local temperature in reactor700 and generate a signal 714 indicative of a detected localtemperature. The signal 714 is transmitted to a control system 720 inany acceptable manner, such as without limitation, fluid coupling,electrical coupling, optical coupling, radiofrequency transmission,acoustic coupling, magnetic coupling, or the like. Responsive to signal714 indicative of the detected local temperature, control system 720determines an appropriate correction (positive or negative) to a localneutronic reactivity of reactor 700 (e.g., to return reactor 700 to adesired operating parameter, such as desired local temperatures duringreactor operations at power). To that end, control system 720 generatesa control signal 724 indicative of a desired correction to localneutronic reactivity. Control signal 724 is transmitted to a dispenser730 of neutron absorbing material. Control signal 724 suitably may betransmitted in the same manner or a different manner as signal 714. Theneutron absorbing material suitably may be any neutron absorbingmaterial as desired for a particular application, such as withoutlimitation Li-6, B-10, or Gd. Dispenser 730 suitably is dispensingmechanism acceptable for a desired application. A reservoir (not shown)may be located locally to dispenser 730 or may be located remotely fromthe dispensing mechanism 730 (e.g., outside a neutron reflector ofreactor 700). Dispenser 730 dispenses the neutron absorbing materialwithin the nuclear fission reactor core responsive to the control signal1124, thereby altering the local neutronic reactivity.

Referring now to FIG. 7B and given by way of non-limiting example,exemplary thermal control may be established with a neutron absorbingfluid. A thermally coupled fluid containing structure 740 contains afluid in thermal communication with a local region of reactor 700. Thefluid in the structure 740 expands or contracts responsive to localtemperature fluctuations. Expansion and/or contraction of the fluid isoperatively communicated to a force coupling structure 750, such aswithout limitation a piston, located external to the nuclear fissionreactor 700. A resultant force communicated by the force couplingstructure 750 is exerted on neutron absorbing fluid, such as Li-6, in aneutron absorbing fluid containing structure 760. The neutron absorbingfluid is dispensed accordingly from the structure 760, thereby alteringthe local neutronic reactivity. In another example, a neutron moderatingfluid may be used instead of, or in addition to, the neutron absorbingfluid. The neutron moderating fluid changes the neutron energy spectrumand lowers the mean neutron energy of the local neutron environment,thereby driving down or lowering neutronic reactivity of nuclear fissionfuel within the nuclear fission reactor 700. In another example, theneutron absorbing fluid and/or the neutron modifying fluid may have amultiple phase composition (e.g., solid pellets within a liquid).

FIG. 7C illustrates details of an exemplary implementation of thearrangement shown in FIG. 7B. Referring now to FIG. 7C, fuel powerdensity in a nuclear fission reactor 701 is continuously regulated bythe collective action of a distributed set of independently-actingthermostating modules, over very large variations in neutron flux,significant variations in neutron spectrum, large changes in fuelcomposition and order-of-magnitude changes in power demand on thereactor. This action provides a large negative temperature coefficientof reactivity just above the design-temperature of reactor 701. Locatedthroughout the fuel-charge in the nuclear fission reactor 701 in a 3-Dlattice (which can form either a uniform or a non-uniform array) whoselocal spacing may be roughly a mean free path of amedian-energy-for-fission neutron (it may be reduced for redundancypurposes), each of these modules includes a pair of compartments, eachone of which is fed by a capillary tube. A small thermostat-bulbcompartment 761 located in the nuclear fission fuel contains a thermallysensitive material, such as without limitation, Li-7, whose neutronabsorption cross-section may be low for neutron energies of interest,while the relatively large compartment 741 is positioned in a differentlocation (e.g., on the wall of a coolant tube) and may contain variableamounts of a neutron absorbing material, such as without limitation,Li-6, which has a comparatively large neutron absorption cross-section.At a pressure of 1 bar, lithium melts at 453K and boils at 1615K, andtherefore is a liquid across typical operating temperature ranges ofreactor 701. As fuel temperature rises, the thermally sensitive materialcontained in the thermostat-bulb 761 expands, and a small fraction of itis expelled (e.g., approximately 10⁻³, for a 100K temperature change inLi-7), potentially under kilobar pressure, into the capillary tube whichterminates on the bottom of a cylinder-and-piston assembly 751 locatedremotely (e.g., outside of the radiation shield) and physically lowerthan the neutron absorbing material's intra-core compartment 741 (in theevent that gravitational forces are to be utilized). There the modestvolume of high-pressure thermally sensitive material drives aswept-volume-multiplying piston in the assembly 751 which pushes alarger (e.g., potentially three order-of-magnitude larger) volume ofneutron absorbing material through a core-threading capillary tube intoan intra-core compartment proximate to the thermostat-bulb which isdriving the flow. There the neutron absorbing material, whose spatialconfiguration is immaterial as long as its smallest dimension is lessthan a neutron mean free path, acts to absorptively depress the localneutron flux, thereby reducing the local fuel power density. When thelocal fuel temperature drops, neutron absorbing material returns to thecylinder-and-piston assembly 751 (e.g., under action of a gravitationalpressure-head), thereby returning the thermally sensitive material tothe thermostat-bulb 761 whose now-lower thermomechanical pressurepermits it to be received.

It will be appreciated that operation of thermostating modules does notrely upon the specific fluids (Li-6 and Li-7) discussed in the aboveexemplary implementation. In one exemplary embodiment, the thermallysensitive material may be chemically, not just isotopically, differentfrom the neutron absorbing material. In another exemplary embodiment,the thermally sensitive material may be isotopically the same as theneutron absorbing material, with the differential neutron absorbingproperties due to a difference in volume of neutronically exposedmaterial, not a difference in material composition.

Reactor Control and Simulation

The aforementioned examples thus demonstrate that fuel, neutronabsorbing material, and other materials may be moved throughout areactor core by several mechanisms with or without moving completeassemblies. Such movements may complicate calculations of nuclideconcentrations (i.e., numbers of atoms and isotopes and nuclear isomersof atoms per unit volume) in the core.

In general, the calculation of nuclide concentrations in the core or anoperating reactor or simulation thereof may be broken into twointerrelated parts: neutron transport and transmutation. Neutrontransport calculations may determine neutron populations (e.g., flux andflux spectrum), while transmutation calculations determine thepopulations of nuclides given a starting population and a neutron flux.

Neutron transport calculations can be done, for example, usingdeterministic methods (e.g., a discrete ordinates method), usingstochastic methods such as a Monte Carlo method, or by using a hybrid ofthe two (e.g., using deterministic methods to calculate certain aspectsin an otherwise Monte Carlo implementation). Deterministic methodstypically solve transport equations using average particle behavior. Adiscrete method typically divides the phase space into many smallvolumes. Neutrons moving between adjacent volumes take a small amount oftime to move a small distance. Thus, calculation approaches theintegro-differential transport equation (having space and timederivatives) as time, volume, and distance are made smaller, i.e.,approach 0.

Monte Carlo methods, on the other hand, obtain answers by simulatingindividual particles and recording some aspects of their averagebehavior. Monte Carlo methods are often used when it is difficult todetermine an example result using a deterministic method. As applied toneutron transport, a Monte Carlo method may simulate the individualprobabilistic events, thus following neutrons through their lifecyclefrom birth to death (e.g., absorption, escape, etc.). The associatedprobability distributions (e.g., represented by continuous and/ordiscrete probability density functions) are randomly sampled todetermine the outcome (e.g., scatter, fission, neutron capture, leakage)at each time step. Collisions may be modeled using physics equations andcross sectional data. The frequency of collisions, and thus neutroninduced reactions such as fission and loss due to absorption by neutronabsorbing materials are, of course, dependent on the concentration offissile isotopes and neutron absorbing materials respectively in thevolume of interest.

Cross-sectional data for an atom represents the effective crosssectional area that an atom presents to a particle for an interaction,e.g., for a neutron, for interactions such as the various scattering andabsorption types. Cross sections typically vary by the atom, theparticle, and the energy of the particle. Thus, a cross section may beused to express the likelihood of a particular interaction of an atomwith an incident particle having a certain energy.

Microscopic properties, such as a microscopic cross section for areaction (e.g., scatter, radiative capture, absorption, fission), areintrinsic properties of a type of nuclei (i.e., of a specific material'snuclei). Macroscopic properties, such as a macroscopic cross section fora reaction, is a property of a volume of the material having aconcentration or density (e.g., in number of atoms per unit volume) ofthe material. Microscopic cross section is typically expressed in unitsof area (e.g., cm² or “barns”—a barn is 10⁻²⁸ m²). Macroscopic crosssections are proportional to the microscopic cross section multiplied bythe density, or equivalently 1/(mean free path length) and thus areexpressed in units of 1/length (e.g., m⁻¹).

Cross sectional data is typically determined by empirical means. Thus,especially for short-lived isotopes, cross sectional data for a largespectrum of neutron energies is simply not available yet. Therefore,performing accurate Monte Carlo calculations on volumes having apopulation of isotopes not having completely known or well-characterizedproperties such as neutron cross-sections can be difficult.Additionally, even if all the cross sectional data for each and everymaterial was well characterized, the computational burden would besignificant. Methods which may help reduce these difficulties and/orcomputational burdens are described in detail elsewhere herein.

Transmutation calculations determine the inventory or concentration ofeach nuclide as it varies, for example, under a neutron flux. Ingeneral, transmutation calculations may be thought of as determining anew population of a material based on the loss rate and the productionrate of the material subject to a given neutron flux. A given atom of amaterial may, for example, fission and produce two fission products;while another atom of the material might be converted to an isotope of alarger atomic mass number (A) after capturing a neutron. Yet anotheratom of the material might beta or alpha decay to another element, andso forth. Thus, the rate of change of an amount of a material in anoperating reactor is typically the sum of the loss rate due to decay,gain rate due to decay, loss due to neutron-induced reactions, and gaindue to neutron-induced reactions.

It is to be appreciated that transmutation calculations for materialsdepend upon the current neutron flux, and neutron flux calculationsdepend upon the current concentration of materials such as fissileisotopes and neutron absorbing materials. These calculations may belinked together in various ways, including but not limited to suchiterative numerical analysis tools such as the Runge-Kutta methods. Acomplete description of Runge-Kutta is not necessary, as it is wellknown in the art. In general, however, explicit Runge-Kutta methods,“solve” the initial value problem

y′=f(t,y),y(t ₀)=y ₀

using the equations

$y_{n + 1} = {y_{n} + {h{\sum\limits_{i = 1}^{5}{b_{i}k_{i}}}}}$where

k ₁ =f(t _(n) ,y _(n)),

k ₂ =f(t _(n) +c ₂ h,y _(n) +a ₂₁ hk ₁),

k ₃ =f(t _(n) +c ₃ h,y _(n) +a ₃₁ hk ₁ +a ₃₂ hk ₂),

k _(s) =f(t _(n) +c _(s) h,y _(n) +a _(s1) hk ₁ +a _(s2) hk ₂ + . . . +a_(s,s-1) hk _(s-1))

To specify a specific Runge Kutta method, one may supply an integer, s,and a set of coefficients a_(ij), b_(ij), and c_(i), The Runge Kuttamethod is consistent if the coefficients are such that:

${{\sum\limits_{j = 1}^{i - 1}a_{ij}} = {{c_{i}\mspace{14mu} {for}\mspace{14mu} i} = 2}},\ldots \mspace{14mu},{s.}$

Thus, for example, a consistent fourth order Runge Kutta is:

y _(n+1) =y _(n)+⅙h(k ₁ +k ₂ +k ₃ +k ₄),

t _(n+1) =t _(n) +h

where

k ₁ =f(t _(n) ,y _(n)),

k ₂ =f(t _(n)+½h,y _(n)+½hk ₁),

k ₃ =f(t _(n)+½h,y _(n)+½hk ₂), and

k ₄ =f(t _(n) +h,y _(n) +hk ₃).

Thus, the next value, y_(n+1), is determined by the present value,y_(n), plus the product of the size of the interval and an estimatedslope. The slope is a weighted average of slopes: k₁ is the slope at thebeginning of the interval, k₂ is the slope at the midpoint of theinterval using slope k₁ to determine the value of y at the pointt_(n)+h/2 using Euler's method; k₃ is again the slope at the midpoint,but now using the slope k₂ to determine the y-value; and k₄ is the slopeat the end of the interval, with its y-value determined using k₃. TheEuler method is a one stage Runge Kutta method. The Euler methodessentially estimates the slope and advances a small step using thatslope. Examples of second order Runge Kutta methods include the midpointmethod and Heun's method.

Thus, an updated amount (e.g., inventory or concentration) of a materialin a reactor core or volume of interest (inside or outside the reactorcore) may be determined by determining an average rate of change of theamount of the material based on the previous amount of the material anda neutron flux. This may be performed individually or simultaneously forall of the materials in the reactor core or the volume of interest. Theneutron flux, in turn, may be determined by determining an average rateof change of flux based on the amount of the materials in the core.

Accuracy of the calculations may be enhanced if subvolumes of a reactorare considered rather than a reactor core in gross. For example, grosscalculations may be performed on a homogenous model of a reactorcore—the core is simulated to have an even distribution of allmaterials. Higher resolution may be obtained by representing the core asa volume comprised of many homogeneous cells, each cell being allowed tohave different concentrations of materials. Although cells need not behomogenous, homogenous cells are typically preferred to simplifycalculations.

If the resolution is high enough, the core may be represented with verygood precision. For example, a three-dimensional geometry of cells, eachhaving a defined geometry and concentrations of materials may be used.Cells may be defined in many ways, including but not limited to by theirbounding surfaces such as equations of surfaces and intersections andunions of regions of space. Transport calculations typically determinefor each cell the number of reactions and boundary crossings to eachneighboring cell.

As illustrated in FIG. 8A, a structure 800 may be formed by cells havingcomplicated shapes. For the sake of simplicity, only two dimensions areshown (i.e., a cross section), but it is understood that cells aretypically three dimensional. Moreover, in this non-limiting example, thelocations and shapes are relatively uniform. For example, exemplary cell802 may be a sphere. Exemplary cell 804 may be a larger sphere excludingthe volume defined by cell 802. Exemplary cell 806 may be a cube,excluding the volume circumscribed by the outer spherical surface ofcell 804. Alternatively, cell 802 could be a cylinder extending somedistance into the figure, cell 804 could be the volume determined by alarger cylinder excluding the volume of cell 802, and cell 806 could bea rectangular prism excluding the volume within the cylinder defined bycell 804's outer surface. In any case, cell 802 may include onecomposition of fuel materials, neutron absorbing materials, andstructural materials. Cell 804 may have a second composition of fuelmaterials, neutron absorbing materials, and structural materials. Cell806 may be a third composition of structural materials only (e.g.,cladding).

As illustrated in FIG. 8B, cells may be combined to form largerstructures. For example, structure 800 may represent a rectangularprism-shaped fuel assembly. Structure 830 includes many structures 800.For example, structure 830 may define a fuel module of six fuel pins byfour fuel pins and fifty fuel pins deep. Thus, even larger structuresmay be formed. For example, as illustrated by FIG. 8C, exemplarystructure 860 may represent a reactor core having an arrangement ofnineteen structures 830 (e.g., fuel modules) each including manystructures 800 (fuel assemblies). Thus, specific physical locations inspace of an actual operating reactor or a detailed reactor design may berepresented by a cell. Calculations may be performed using a detailedmodel representing an actual reactor during operation. The results maybe used to make decisions regarding reactor control. Similarly,calculations may be performed on a representation of a proposed reactorto test operating procedures or to test proposed fuel and neutronabsorbing material loading.

Transmutation and transport calculations may be performed for each cell.For a complex model, this can result in a large computational burden duein part to the large number of cells. The computational burden is alsoincreased by the number of materials which may be present in each cell.Prior to operation, a reactor already contains a large number ofmaterials (e.g., various fuel isotopes, installed neutron absorbingmaterial, structural isotopes, moderator, reflectors, etc.). Immediatelyupon operation, however, the number of materials (e.g., isotopes) in thereactor increases significantly due to neutron capture and especiallyneutron-induced fission.

The distribution of fission products from a fission of a given isotopeinduced by a neutron of a given energy may be described by a fissionproduct yield curve. FIG. 9 illustrates an exemplary fission productyield curve 900. It should be appreciated that the graph illustrates thetotal fission yield in percent of fission products having each massnumber (A). More than one isotope may have a given mass number. Thus,fission products having a mass number of, for example, 140, fall underthe point on the curve defined by mass number=140. In this example, thefission products produced by the thermal fission of U-235 is illustratedon fission product yield curve 900. Curves for fissions of U-235 inducedby fast neutrons will have a similar but different shape. Neutronenergies may be classified in more detail than “fast” or “thermal.”Also, fission yield curves for other fissile isotopes will have asimilar but different shape. In general, however, fission yield curvesfollow this “M” shape having two peaked “humps.” Thus, the curve may bedivided into two portions, left curve portion 912 which includes a leftpeak 922, and right curve portion 914 which includes a right peak 924.Thus area 902 falls under left peak 922 and left curve portion 912 andarea 904 falls under right peak 924 and right curve portion 914. As areactor operates, the level of fission products tends to increase due tofission (i.e., have a production rate due to fission), but tends todecrease due to decay and neutron capture or “burnout” (i.e., have lossrates due to decay and capture). Transmutation calculations may be usedto determine or approximate these levels during reactor operation.

As discussed elsewhere herein, reactor control systems, such as controlsystem 720, may determine appropriate corrections (positive or negative)to a local neutronic reactivity of reactor 700 (e.g., to return reactor700 to a desired operating parameter, such as desired local temperaturesduring reactor operations at power). To that end, control systems maygenerate a control signal (e.g., control signal 724) indicative of adesired correction to local neutronic reactivity. Reactor controlsystems and control signals are not limited to the embodiments such ascontrol system 720 and control signal 724. Reactor Control Systems mayalso control other neutron affecting or absorbing features such ascontrol rods, to control and/or shut down the reactor as desired, whichis well known in the art. Reactor Control Systems may also generatecontrol signals to order changes in various flows, e.g., the flow ofheat absorbing material (e.g., coolant) through the reactor or portionsof the reactor by ordering changes in reactor coolant pump (e.g.,reactor coolant pumps 22) operation and/or various valve positions inthe reactor system, including but not limited to reactor closures (e.g.,closures 24) or reactor coolant shutoff valves, steam shutoff valves,etc. Reactor Control Systems may also order changes in breaker positions(e.g., reactor coolant pump power supply breakers, steamturbine-generator output breakers, etc.). As is well known in the art,Reactor Control Systems may have temperature inputs (e.g., controlsystem 720 receiving input from temperature detectors 710) in additionto neutron detectors (e.g., to sense neutron flux to determine reactorpower or local reactor power at a portion of the core), and flow andposition detectors (e.g., venturi-type flow detectors, valve positionindicators, breaker position indicators). Thus, Reactor Control Systemsmay control the flow of heat absorbing material (e.g., coolant) throughthe reactor and/or portions of the reactor to control overalltemperatures and local temperatures in response to overall reactorthermal power and/or local reactor thermal power. Reactor ControlSystems may also provide operator indications and accept operatorinputs. Thus, a Reactor Control System monitors reactor operations, mayprovide some automatic control features (such as changing flow rates andmoving control rods or otherwise positioning neutron affecting orabsorbing materials, which are described in more detail elsewhereherein), displays operational parameters, and accepts and executesoperator inputs for manual control actions.

Example Computer System

Some aspects and/or features of the disclosed subject matter can beimplemented by software, firmware, hardware, or a combination thereof.Calculations may be approximated using table look-ups. Hardwareimplementations of individual components are not limited to digitalimplementations and may be analog electrical circuits. Additionally,embodiments may be realized in a centralized fashion in at least onecommunication system, or in a distributed fashion where differentelements may be spread across several interconnected communicationsystems. Any kind of computer system or other apparatus adapted forcarrying out the methods described herein may be suited.

FIG. 10 illustrates an example computer system 1000 in which the presentsubject matter, or portions thereof, can be implemented ascomputer-readable code. Various embodiments are described in terms ofthis example computer system 1000. After reading this description, itwill become apparent to a person skilled in the relevant art how toimplement the disclosed subject matter using other computer systemsand/or computer architectures.

Computer system 1000 includes one or more processors, such as processor1004. Processor 1004 can be a special purpose or a general purposeprocessor. Processor 1004 is connected to a communication infrastructure1006 (for example, a bus or network).

Computer system 1000 also includes a main memory 1008, preferably randomaccess memory (RAM), and may also include a secondary memory 1010.Secondary memory 1010 may include, for example, a hard disk drive 1012,a removable storage drive 1014, any type of non-volatile memory, and/ora memory stick. Removable storage drive 1014 may comprise a floppy diskdrive, a magnetic tape drive, an optical disk drive, a flash memory, orthe like. The removable storage drive 1014 reads from and/or writes to aremovable storage unit 1018 in a well known manner. Removable storageunit 1018 may comprise a floppy disk, magnetic tape, optical disk, etc.which is read by and written to by removable storage drive 1014. As willbe appreciated by persons skilled in the relevant art(s), removablestorage unit 1018 includes a computer usable storage medium havingstored therein computer software and/or data.

In alternative implementations, secondary memory 1010 may include othersimilar means for allowing computer programs or other instructions to beloaded into computer system 1000. Such means may include, for example, aremovable storage unit 1022 and an interface 1020. Examples of suchmeans may include a program cartridge and cartridge interface (such asthat found in video game devices), a removable memory chip (such as anEPROM, or PROM) and associated socket, and other removable storage units1022 and interfaces 1020 which allow software and data to be transferredfrom the removable storage unit 1022 to computer system 1000.

Computer system 1000 may also include a communications interface 1024.Communications interface 1024 allows software and data to be transferredbetween computer system 1000 and external devices. Communicationsinterface 1024 may include a modem, a network interface (such as anEthernet card), a communications port, a PCMCIA slot and card, or thelike. Software and data transferred via communications interface 1024are in the form of signals which may be electronic, electromagnetic,optical, or other signals capable of being received by communicationsinterface 1024. These signals are provided to communications interface1024 via a communications path 1026. Communications path 1026 carriessignals and may be implemented using wire or cable, fiber optics, aphone line, a cellular phone link, an RF link or other communicationschannels.

Computer system 1000 may also be coupled to a Reactor Control system1030. Reactor Control System 1030 may be directly interfaced to thecommunications infrastructure 1006 as shown in the figure. ReactorControl System may also be interfaced via communications interface 1024or communications interface 1024 and communications path 1026.

In this document, the terms “computer program medium” and “computerusable medium” are used to generally refer to media such as removablestorage unit 1018, removable storage unit 1022, and a hard diskinstalled in hard disk drive 1012. Signals stored elsewhere and carriedover communications path 1026 can also embody the logic describedherein. Computer program medium and computer usable medium can alsorefer to memories, such as main memory 1008 and secondary memory 1010,which can be memory semiconductors (e.g. DRAMs, etc.). These computerprogram products are means for providing software to computer system1000.

Computer programs (also called computer control logic) are stored inmain memory 1008 and/or secondary memory 1010. Computer programs mayalso be received via communications interface 1024. Such computerprograms, when executed, enable computer system 1000 to implement thepresent subject matter as discussed herein. In particular, the computerprograms, when executed, enable processor 1004 to be used in theperformance of processes of the present subject matter, such as themethods illustrated by the flowcharts described elsewhere herein.Accordingly, such computer programs represent controllers of thecomputer system 1000. Where the disclosed subject matter is implementedusing software, the software may be stored in a computer program productand loaded into computer system 1000 using removable storage drive 1014,interface 1020, hard drive 1012 or communications interface 1024.

The present subject matter is also directed to computer program productscomprising software stored on any computer useable medium. Computerprograms or software in the present context means any expression, in anylanguage, code or notation, of a set of instructions intended to cause asystem having an information processing capability to perform aparticular function either directly or after either or both of thefollowing: a) conversion to another language, code or notation; b)reproduction in a different material form. Such software, when executedin one or more data processing device, causes a data processingdevice(s) to operate as described herein. Embodiments employ anycomputer useable or readable medium, known now or in the future.Examples of computer useable mediums include, but are not limited to,primary storage devices (e.g., any type of random access memory),secondary storage devices (e.g., hard drives, floppy disks, CD ROMS, ZIPdisks, tapes, magnetic storage devices, optical storage devices, MEMS,nanotechnological storage device, etc.), and communication mediums(e.g., wired and wireless communications networks, local area networks,wide area networks, intranets, etc.).

Methods for Mapping Reactor Materials

Now that illustrative embodiments of nuclear reactors and reactorcontrol and simulation have been discussed, illustrative methodsassociated therewith will now be discussed.

Following are a series of flowcharts depicting implementations ofprocesses. For ease of understanding, the flowcharts are organized suchthat the initial flowcharts present implementations via an overall “bigpicture” viewpoint and thereafter the following flowcharts presentalternate implementations and/or expansions of the “big picture”flowcharts as either sub-steps or additional steps building on one ormore earlier-presented flowcharts. Those having skill in the art willappreciate that the style of presentation utilized herein (e.g.,beginning with a presentation of a flowchart(s) presenting an overallview and thereafter providing additions to and/or further details insubsequent flowcharts) generally allows for a rapid and easyunderstanding of the various process implementations. In addition, thoseskilled in the art will further appreciate that the style ofpresentation used herein also lends itself well to modular designparadigms. The blocks may be performed in any order or concurrentlyunless specified otherwise. Some embodiments do not require theperformance of each and every block, regardless whether the block orblocks is/are explicitly labeled or described as optional. Otherembodiments require the repetition of one or more blocks, regardlesswhether the block is labeled or described as repeated.

Referring now to FIG. 11, an illustrative method 1100 is provided forsimulating and/or controlling a nuclear reactor. The method 1100 startsat a block 1105.

At block 1105, a flux in a first cell is determined based on theamount(s) of at least one material in at least the first cell. The fluxdetermination may be further based on the amounts of more than onematerial in the first cell and/or a previous flux in the first cell.Also, the flux determination may be further based on the amounts one ormore materials in one or more other cells. For example, a flux may bedetermined by a transport calculation (e.g., solving neutron transportequations). A “flux” may be any flux (e.g., photon, alpha, beta, etc.),but is typically a neutron flux. The flux may be determined by numericalanalysis methods using an average rate of change of the flux. Theaverage rate of change of the flux may be a weighted average (e.g., asdetermined by a Runge Kutta method or any other method). The flux may bedependent upon the amount(s) of one or more materials in the first cell.The flux may be further dependent upon the amount(s) of one or morematerials in one or more additional cells. An “amount” may be a mass ora number (e.g., number of atoms) or may be a density/concentration(e.g., mass or number of particles per unit volume). A cell represents aphysical location or region in a nuclear reactor. The reactor may be,for example, real or simulated, currently operating, or under design.The reactor may be any type or sub-type of reactor, including lightwater reactor, heavy water reactor, pressurized water reactor, boilingwater reactor, propagating nuclear fission deflagration wave reactor,etc. The reactor is typically represented by many homogeneous cells, butheterogeneous cells may be used. Each cell may have the same ordifferent shape or volume as any other cell. A material may be one ormore of any element, molecule, family of elements, family of molecules,isotope, family of isotopes, isomers of isotopes, fertile isotope(s),fission product(s), fission product poisons, etc. Materials aretypically elements and isotopes of elements. Thus, U-235 and U-238 arethus typically two different materials.

At block 1110, the average rate(s) of change of the one or moreamount(s) of one or more material(s) in the first cell is determinedbased on previous amount(s) of the material(s) and a flux in the firstcell. For example, average rates of change may be determined by atransmutation rate calculation. The average rate of change of the amountof one or more materials may be determined by numerical analysis methodsusing an average rate of change of the amount. The average rate ofchange of the flux may be a weighted average (e.g., as determined by aRunge Kutta method or any other method). The amount may be dependentupon the flux in the first cell. The average rates of change for the oneor more materials may be solved individually or simultaneously (such aswhen coupled through transmutation equations).

At block 1115, updated amount(s) in the first cell for the material(s)are determined based on the average rate(s) of change. For example,updated amounts may be determined by performing transmutationcalculations. The updated amounts for the one or more materials may besolved individually or simultaneously (such as when coupled throughtransmutation equations).

At block 1120, at least one move quantity is determined. A move quantitymay be any quantity of one or more materials such as a quantity of amaterial that is desired to be moved into or out of a cell. In thisblock one or more move quantities may each apply to one or morematerials in the first cell. A move quantity may be determined inresponse to one or more reactor parameters such as a flux or a fluence,a power level (local or overall), a temperature, etc. A reactorparameter may be compared to a threshold or set point for thatparameter. This block may be repeated as suitable, e.g., for each of oneor more materials in the first cell.

At block 1125, the updated amount(s) in the first cell is (are) adjustedby the move quantity(-ies). One or more move quantities are each appliedto the amounts of one or more materials in the first cell, thusincreasing or decreasing each affected amount. A move amount of zero maybe used to signify no change. In an embodiment, a material may be movedoutside the reactor. In this case, blocks 1130 through 1145 may beskipped.

At block 1130, a flux in a second cell is determined based on amount(s)of at least one material in the second cell. As discussed above, theflux determination may be further based on the amounts of more than onematerial in the second cell. Also, the flux determination may be furtherbased on the amounts one or more materials in one or more other cells.

At block 1135, average rate(s) of change of the amount(s) of thematerial(s) in the second cell based on previous amount(s) of thematerial(s) and a flux in the second cell is determined.

At block 1140, updated amount(s) in the second cell for the material(s)is (are) determined based on the average rate(s) of change in the secondcell.

At block 1145, the updated amount(s) in the second cell is (are)adjusted by the move quantity(-ies).

At block 1150, a control action for a nuclear reactor is determined. Acontrol action may be a change (positive or negative) to a localneutronic reactivity of a reactor using any neutron affecting orabsorbing features such as movement of neutron absorbing materials orfluids, control rods, etc.; a change in one or more various flows, e.g.,the flow of heat absorbing material (e.g., coolant) through the reactorby ordering changes in reactor coolant pump operation and/or variousvalve positions in the reactor system, including but not limited toreactor closures or reactor coolant shutoff valves, steam shutoffvalves, etc.; a change in one or more breaker positions (e.g., reactorcoolant pump power supply breakers, steam turbine-generator outputbreakers, etc.); or the like. The determined control action may bedisplayed to a user. In an embodiment, this block is optional.

At block 1155, a control action for the nuclear reactor is performed.This performance may be automatic or manual. In an embodiment, thisblock is optional.

At block 1160, approximately the move quantity(-ies) of the material(s)is/are transferred to/from the first cell location in a reactor. In thisblock, an actual of amount of at least one substance corresponding toone or more of the at least one material is transferred from or to onelocation (i.e., from the first cell location or to the first celllocation). This block may be performed in conjunction with block 1155 orseparately. The transferred amount of a substance (i.e., approximatelythe move quantity of the corresponding material or materials) may, butis not required, to be associated with a component in the locationrepresented by the cell (e.g., an assembly including fuel, neutronabsorbing material, structural components, or any combination of these).In an embodiment such as a simulation or evaluation of a reactor design,this step is optional.

The method stops at block 1160, but may continue to point A as indicatedin other methods in other figures.

Referring now to FIG. 12, an illustrative method 1200 is provided forsimulating and/or controlling a nuclear reactor. The method 1200 startsat a block 1210. As illustrated by point A, method 1200 may be precededby method 1100.

At block 1210, approximately the move quantity(-ies) of the material(s)is/are transferred from/to a second cell. In this block, an actualamount of at least one substance corresponding to one or more of the atleast one material is transferred from or to the location of the second(i.e., from the second cell location or to the second cell location).For example, in conjunction with block 1160 of method 1100, a quantityof a substance, approximately equal to the determined move quantity orquantities of the corresponding materials, may be transferred from thefirst cell to the second cell or vice versa.

The method stops at block 1210.

Referring now to FIG. 13, an illustrative method 1300 is provided forsimulating and/or controlling a nuclear reactor. The method 1300 startsat a block 1305. As illustrated by point A, method 1300 may be precededby method 1100. Illustrative method 1300 provides an exemplary method ofmoving matter in a four-cell loop through the reactor. At each cell inthe loop, the amount and type of matter moved need not be identical. Aperson of skill in the art would understand that the four-cell loop maybe expanded or contracted as suitable (i.e., include fewer or morecells).

At block 1305, approximately the move quantity(-ies) of material(s)is/are transferred to a second cell. For example, in conjunction withblock 1160 of method 1100, a quantity of a substance, approximatelyequal to the determined move quantity or quantities of the correspondingmaterials, may be transferred from the first cell to the second cell.

At block 1310, second move quantity(-ies) is (are) determined. Thesecond move quantity or quantities may be calculated in any way asdescribed above.

At block 1315, amount(s) in the second cell is (are) adjusted by thesecond move quantity(-ies).

At block 1320, approximately the second move quantity(-ies) ofmaterial(s) are transferred from the second cell to a third cell.

At block 1325, third move quantity(-ies) is (are) determined.

At block 1330, amount(s) in the third cell is (are) adjusted by thethird move quantity(-ies).

At block 1335, approximately the third move quantity(-ies) ofmaterial(s) is (are) transferred from the third cell to a fourth cell.

At block 1340, fourth move quantity(-ies) is (are) determined.

At block 1345, amount(s) in the fourth cell is (are) adjusted by thefourth move quantity(-ies).

At block 1350, approximately the fourth move quantity(-ies) ofmaterial(s) is (are) transferred from the fourth cell to a first cell

The method stops at block 1350.

Referring now to FIG. 14, an illustrative method 1400 is provided forsimulating and/or controlling a nuclear reactor. The method 1400 startsat a block 1405. As illustrated by point A, method 1400 may be precededby method 1100. Illustrative method 1400 illustrates, inter alia, mixingquantities of one or materials from a first cell and a third cell, andtransferring at least a portion of the mixture back to the first cell.One or more additional iterations of neutron flux and transmutationcalculations may optionally occur. A person of skill in the art wouldunderstand that this illustrative method could be expanded or contractedto include various mixing methods using fewer or more cells.

At block 1405, approximately the move quantity(-ies) of materials is(are) transferred to a second cell.

At block 1410, second move quantity(-ies) is (are) determined.

At block 1415, amount(s) in the second and third cells is (are) adjustedby the second move quantity(-ies).

At block 1420, approximately the second move quantity(-ies) aretransferred to/from a third cell and from/to the second cell.

At block 1425, a new average rate(s) of change of the amount(s) of thematerial(s) in the first cell based on current amount(s) of thematerial(s) and an updated flux in the first cell is (are) determined.

At block 1430, new updated amount(s) in the first cell for thematerial(s) is (are) determined based on the new average rate(s) ofchange.

At block 1435, third move quantity (-ies) is (are) determined.

At block 1440, amount(s) in the second and first cell is (are) adjustedby the third move quantity(-ies).

At block 1445, approximately the third move quantity (-ies) is (are)transferred from the second cell to the first cell.

The method stops at block 1445.

Referring now to FIG. 15, an illustrative method 1500 is provided forsimulating and/or controlling a nuclear reactor. As illustrated by pointA, method 1500 may be preceded by method 1100. Illustrative method 1500provides an exemplary method of, inter alia, transferring quantities ofone or more materials to a cell location for holding (e.g., a holdingtank or reservoir, etc.). While the material is in the holding celllocation, one or more additional iterations of neutron flux andtransmutation calculations may optionally occur. Material may also betransferred out of the holding cell (e.g., to a location that is notrepresented by the first cell). A person of skill in the art wouldunderstand that this illustrative method could be expanded or contractedto include various holding methods using fewer or more cells. The method1500 starts at a block 1505.

At block 1505, approximately the move quantity(-ies) of material(s) is(are) transferred to a second cell.

At block 1510, second move quantity(-ies) is (are) determined.

At block 1515, amount(s) in the second cell is (are) adjusted by thesecond move quantity(-ies).

At block 1520, approximately the second move quantity(-ies) ofmaterial(s) is (are) transferred from the second cell.

At block 1525, new average rate(s) of change of the amount(s) of thematerial(s) in the first cell is (are) determined based on currentamount(s) of the material(s) and an updated flux in the first cell.

At block 1530, new updated amount(s) in the first cell for thematerial(s) is (are) determined based on the new average rate(s) ofchange.

At block 1535, an average rate(s) of change of the amount(s) of thematerial(s) in the second cell is (are) determined based on currentamount(s) of the material(s) and a flux in the second cell.

At block 1540, updated amount(s) in the second cell for the material(s)is (are) determined based on the average rate of change in the secondcell.

At block 1545, third move quantity(-ies) is (are) determined.

At block 1550, amount(s) in the first and second cells is (are) adjustedby the third move quantity(-ies).

At block 1555, approximately the third move quantity(-ies) ofmaterial(s) is (are) transferred from the second cell to the first cell.

The method stops at block 1555.

Referring now to FIG. 16, an illustrative method 1600 is provided forsimulating and/or controlling a nuclear reactor. Illustrative method1600 differs from illustrative method 1110, but some steps may besimilar. For example, illustrative method 1600 provides an example of acontinuous move of material(s) rather than discrete moves. The method1600 starts at a block 1605.

At block 1605, a flux in a first cell is determined based on amount(s)of at least one material in the first cell. As discussed above, the fluxdetermination may be further based on the amounts of more than onematerial in the first cell. Also, the flux determination may be furtherbased on the amounts one or more materials in one or more other cells.This block is similar to block 1105.

At block 1610, average rate(s) of change of the amount(s) of thematerial(s) in the first cell is (are) determined based on previousamount(s) of the material(s) and a flux in the first cell. This block issimilar to block 1110.

At block 1615, at least one move rate for the material(s) in the firstcell is (are) determined. A move rate may be any rate of movement of oneor more materials such as a quantity of a material that is desired to bemoved into or out of a cell. In this block one or more move rates mayeach apply to one or more materials in the first cell. A move rate maybe determined in response to one or more reactor parameters such as aflux or a fluence, a power level (local or overall), a temperature, etc.A reactor parameter may be compared to a threshold or set point for thatparameter. This block may be repeated as suitable, e.g., for each of oneor more materials in the first cell.

At block 1620, the average rate(s) of change in the first cell is (are)adjusted based on the move rate(s) for the material(s) in the firstcell. For example, an average rate of change of a material in the firstcell may be adjusted (increased or decreased) based on a determined moverate. The adjustment may be made to a single average rate of change orto individual rates of change which are averaged (e.g., in a straightaverage or a weighted average). The average rates of change for morethan one material may be solved individually or simultaneously (such aswhen coupled through transmutation equations).

At block 1625, updated amount(s) in the first cell is (are) determinedbased on the adjusted average rate(s) of change. In an embodiment, amaterial may be moved outside the reactor. In this case, blocks 1630through 1645 may be skipped.

At block 1630, a flux in a second cell is determined based on amount(s)of at least one material in the second cell. As discussed above, theflux determination may be further based on the amounts of more than onematerial in the first cell. Also, the flux determination may be furtherbased on the amounts one or more materials in one or more other cells.

At block 1635, average rate(s) of change of the amount(s) of thematerial(s) in the second cell is (are) determined based on previousamount(s) of the material(s) and a flux in the second cell.

At block 1640, the average rate(s) of change in the second cell is (are)determined based on move rate(s) of at least one material.

At block 1645, updated amount(s) in the second cell is (are) based onthe adjusted average rate(s) of change.

At block 1650, a control action for a nuclear reactor is determined. Inan embodiment, this block is optional.

At block 1655, a control action for nuclear reactor is performed. In anembodiment, this block is optional.

At block 1660, material(s) is (are) transferred to/from the first cellat a transfer rate approximately equivalent to the move rate(s) of thematerial(s). As with block 1160, an actual amount of at least onesubstance is transferred, but the transfer is at a transfer rateequivalent to the appropriate move rate. In an embodiment such as asimulation or evaluation of a reactor design, this step is optional.

The method stops at block 1660, but may continue to point B as indicatedin other methods in other figures.

Referring now to FIG. 17, an illustrative method 1700 is provided forsimulating and/or controlling a nuclear reactor. The method 1700 startsat a block 1705. As illustrated by point B, method 1700 may be precededby method 1600.

At block 1705, material(s) is (are) transferred from/to a second cell atthe transfer rate(s) approximately equivalent to the move rate(s) of thematerials. In this block, an actual of amount of at least one substancecorresponding to one or more of the at least one material is transferredfrom or to the location of the second (i.e., from the second celllocation or to the second cell location) at the move rate(s). Forexample, in conjunction with block 1660 of method 1600, a quantity of asubstance, approximately equal to the determined move quantity orquantities of the corresponding materials, may be transferred from thefirst cell to the second cell or vice versa at the appropriate moverate(s).

The method stops at block 1705.

Referring now to FIG. 18, an illustrative method 1800 is provided forsimulating and/or controlling a nuclear reactor. The method 1800 startsat a block 1805. As illustrated by point B, method 1800 may be precededby method 1600. Illustrative method 1800 provides an exemplary method ofmoving matter at various rates in a four-cell loop through the reactor.At each cell in the loop, the rate and type of matter moved need not beidentical. A person of skill in the art would understand that thefour-cell loop may be expanded or contracted as suitable (i.e., includefewer or more cells).

At block 1805, material(s) is (are) transferred to the second cell attransfer rate(s) approximately equivalent to the move rate(s) of thematerials.

At block 1810, a second (set of) move rate(s) for material(s) in thesecond cell is (are) determined.

At block 1815, further adjustment(s) to average rate(s) of change in thesecond cell is (are) made based on the second (set of) move rate(s) ofthe material(s).

At block 1820, material(s) is (are) transferred from the second cell toa third cell at approximately the second (set of) move rate(s).

At block 1825, move rate(s) for material(s) in the third cell is (are)determined.

At block 1830, average rate(s) of change for material(s) in the thirdcell is (are) adjusted by the second (set of) move rates for the secondcell and/or the determined move rates of the third cell.

At block 1835, material(s) is (are) transferred from the third cell to afourth cell at approximately the determined move rate(s) of the thirdcell.

At block 1840, move rate(s) for material(s) in the fourth cell is (are)determined.

At block 1845, average rate(s) of change for material(s) in the fourthcell is (are) adjusted by the determined move rate(s) of the third celland/or the determined move rates of the fourth cell.

At block 1850, material(s) is (are) transferred from the fourth cell tothe first cell at approximately the determined move rates of the fourthcell.

The method stops at block 1850.

Referring now to FIG. 19, an illustrative method 1900 is provided forsimulating and/or controlling a nuclear reactor. Illustrative method1900 illustrates, inter alia, mixing quantities of one or materials froma first cell and a third cell, and transferring at least a portion ofthe mixture back to the first cell. The transfers occur at variousrates. One or more additional iterations of neutron flux andtransmutation calculations may optionally occur during the transfer. Aperson of skill in the art would understand that this illustrativemethod could be expanded or contracted to include various mixing methodsusing fewer or more cells. The method 1900 starts at a block 1905. Asillustrated by point B, method 1900 may be preceded by method 1600.

At block 1905, material(s) is (are) transferred to the second cell attransfer rate(s) approximately equivalent to the move rate(s) of thematerials.

At block 1910, a second (set of) move rate(s) for material(s) of thesecond cell is (are) determined.

At block 1915, further adjustments are made to average rate(s) of changein the second cell and to the average rate(s) of change in a third cellbased on the second (set of) move rate(s) of the material(s).

At block 1920, material(s) is (are) transferred at approximately thesecond (set of) move rate(s) to/from a third cell from/to the secondcell.

At block 1925, new average rate(s) of change of the amount(s) of thematerial(s) in the first cell is (are) determined based on currentamount(s) of the material(s) and an updated flux in the first cell.

At block 1930, new updated amount(s) in the first cell for thematerial(s) is (are) determined based on the new average rate(s) ofchange.

At block 1935, a third (set of) move rate(s) is determined formaterial(s) of the second cell.

At block 1940, average rate(s) of change of material(s) in the first andsecond cells is (are) adjusted using the third (set of) move rate(s).

At block 1945, material(s) is (are) are transferred at approximately thethird (set of) move rate(s) from the second cell to the first cell.

The method stops at block 1945.

Referring now to FIG. 20, an illustrative method 2000 is provided forsimulating and/or controlling a nuclear reactor. Illustrative method2000 provides an exemplary method of, inter alia, transferringquantities of one or more materials to a cell location for holding(e.g., a holding tank or reservoir, etc.) at various rates. While thematerial is being transferred to/from the holding cell location, one ormore additional iterations of neutron flux and transmutationcalculations may optionally occur. Material may also be transferred outof the holding cell (e.g., to a location that is not represented by thefirst cell). The method 2000 starts at a block 2005. As illustrated bypoint B, method 2000 may be preceded by method 1600.

At block 2005, material(s) is (are) transferred to the second cell attransfer rate(s) approximately equivalent to the move rate(s) of thematerials.

At block 2010, a second (set of) move rate(s) for material(s) of thesecond cell is determined.

At block 2015, further adjustment(s) is (are) made to average rate(s) ofchange in the second cell based on the second (set of) move rate(s) ofthe material(s).

At block 2020, material(s) is (are) transferred at approximately thesecond (set of) move rate(s) from the second cell.

At block 2025, new average rate(s) of change of the amount(s) of thematerial(s) in the first cell is (are) determined based on currentamount(s) of the material(s) and an updated flux in the first cell.

At block 2030, new updated amount(s) in the first cell is (are)determined for the material(s) based on the new average rate(s) ofchange.

At block 2035, new average rate(s) of change of the amount(s) of thematerial(s) in the second cell is (are) determined based on currentamount(s) of the material(s) and a flux in the second cell.

At block 2040, a third (set of) move rate(s) for materials in the secondcell is determined.

At block 2045, the average rate(s) of change of the amount(s) ofmaterial(s) in the first cell is (are) determined based on the third(set of) move rate(s) for materials in the second cell.

At block 2050, the average rate(s) of change of the amount(s) ofmaterial(s) in the second cell is (are) determined based on the third(set of) move rate(s) for materials in the second cell.

At block 2055, material(s) is (are) transferred at approximately thethird (set of) move rate(s) from the second cell to the first cell.

The method stops at block 2055.

Referring now to FIG. 21, an illustrative method 2100 is provided forsimulating andLor controlling a nuclear reactor. As discussed elsewhereherein, neutron transport equations may use cross sectional data forsome or all of the materials in the reactor. Method 2100 illustrates anon-limiting example method that may have one or more of the followingbenefit. It may reduce the computational burden, reduce the need forexhaustive cross sectional data for each reaction for each targetparticle across a spectrum of incident particle energies, and/or improveaccuracy of current methods. The method 2100 starts at a block 2105.

At block 2105, a neighbor to a principal material in a first cell isselected. A principal material is a material of one or more materials ina reactor or reactor cell that may be represented by an agent material.In an embodiment, principal materials may be fission products (elements,isotopes, and/or isomers of isotopes). A principal material might not bewell-characterized with respect to a microscopic property such asmicroscopic cross sectional data. For example, some of the crosssections for scattering, radiative capture, fission, etc. reactions withneutrons of various energies may not be known. Also, the principal'sinformation may be well-known, but for other reasons (such as reducingcomputational burden), the principal material may be represented by aneighbor material which will act as an agent. The neighbor may beselected from a plurality of agent materials. In an embodiment, multipleneighbors may be selected from the plurality of agent materials torepresent the principal material as agents for more than one property.Agent materials are typically well-characterized with regard topertinent microscopic cross sectional data. In an embodiment, agentmaterials are actual materials (e.g, isotopes) with empiricallydetermined microscopic quantities. In a further embodiment, agentmaterials include one or more fictional materials. A fictional materialis essentially a collection of fictional values for various crosssections and optionally other properties. A neighbor may be chosen toact as the agent for the principal on one or more of many criteria.Typically, a neighbor has microscopic properties such that a certaindensity of the neighbor will have similar macroscopic properties as theexisting density of the principal. Thus, a neighbor may be selected toact as an agent for the principal based on a comparison of microscopicproperties of the principal to the microscopic properties of each of theneighbors. Microscopic properties may be approximated having one or morediscrete values with respect to incident particle (e.g., neutron)energy, or may be evaluated as a function of incident particle energy.In an embodiment, the selection of a neighbor or neighbors is limited toa selection from one or more agent materials that are also fissionproducts of the fissions of one or more fissile materials. The fissionproducts may further be limited to fissions induced by neutrons and/orneutrons of certain energy levels. In an embodiment further limiting theselection, potential neighbors may be chosen from agent materials underthe same “hump” as the principal material of a fission yield curve(e.g., left curve portion 912 or right curve portion 914 of fissionyield curve 900 illustrated in FIG. 9). In an embodiment, the number ofpotential principal materials is larger than the number of agentmaterials. For example, the known fission product isotopes number in thethousands. In a further embodiment, the number of agent materials islimited to a relatively small number (e.g., under 100, 50, 30, or 20).In an embodiment, the number of agent materials is limited to 12. In anembodiment, this block is performed by exemplary method 2200 describedbelow.

At block 2110, a proxy amount of the selected neighbor or neighbors isdetermined. As discussed above, a neighbor might have microscopicproperties such that a certain density of the neighbor will have similarmacroscopic properties as the existing density of the principal. Theproxy amount is the amount or density (e.g., concentration) of theneighbor that will serve to represent the principal in a givenconcentration. As with microscopic properties, macroscopic propertiesmay be approximated as one or more discrete values or as a function ofincident particle energy.

At block 2115, blocks 2105 and 2110 are repeated for each of a pluralityof principal materials in the first cell. In this block, a neighbor orneighbors is/are selected to act as an agent(s) for each of theplurality of principal materials (which may or may not make up all ofthe materials in the cell). A given agent material may be selected as aneighbor for more than one principal material. Other agent materials maynot be selected to be any principal's neighbor. Proxy amounts of eachagent are determined for each principal to which the agent is aneighbor.

At block 2120, a summed proxy amount for each agent material isdetermined. In this block, a total proxy amount of each agent isdetermined based on the proxy amounts for each neighbor of the agentmaterial. For example, suppose an agent material was selected to be theneighbor of three different principal materials. After performance ofthe previous blocks, the agent may have three proxy amounts (one foreach principal). In this block, a summed proxy amount is determinedbased on the three proxy amounts (e.g., by summing them).

At block 2125, a flux in the first cell is determined based on thesummed proxy amounts of each agent material the first cell. As describedelsewhere herein, for example, a flux may be determined by a transportcalculation (e.g., solving neutron transport equations to determine aneutron flux) and may be further based on the summed proxy amounts ofeach agent material in one or more other cells. The flux may beapproximated by one or more discrete values, or may be a continuousfunction, thus describing a flux spectrum. The flux may be space and/orenergy dependent. The flux may be determined by numerical analysismethods including Monte Carlo methods. The average rate of change of theflux may be a weighted average (e.g., as determined by a Runge Kuttamethod or any other method). The flux may be dependent upon the amountof each of one or more materials in the first cell. Instead of using theactual amounts (e.g., concentrations) of each material in the cell, thecalculation instead uses the summed proxy amounts of each agentmaterial. Thus, the cross sectional data and concentrations of theprincipal materials are accounted for in a flux determination (e.g.,neutron transport calculations) by agents having proxy concentrations.In embodiments where the number of agents is relatively small, thecomputational burden may be reduced significantly.

At block 2130, an updated amount of one or more materials (principal oragent) is determined based on the previous amount of the materials andthe flux (e.g., the estimated average flux) in the first cell. The oneor more materials may be a subset of the materials in the cell. Forexample, updated amounts may be determined by a transmutationcalculation, which may take into account production rates (e.g., basedon reactions rates such as fission rates) and decay rates (e.g., usingdecay constants). The updated amounts for the one or more materials maybe solved individually or simultaneously (such as when coupled throughtransmutation equations). The calculation may include calculating theupdated amount based on a specified length of time.

At block 2135, a control action for a nuclear reactor is determined. Asdescribed above, the control action may be a change (positive ornegative) to a local neutronic reactivity of a reactor using any neutronaffecting or absorbing features such as movement of neutron absorbingmaterials or fluids, control rods, etc.; a change in one or more variousflow rates for any reason including but not limited to localized oroverall reactor power, e.g., the flow of heat absorbing material (e.g.,coolant) through the reactor or portions of the reactor by orderingchanges in reactor coolant pump operation and/or various valve positionsin the reactor system, including but not limited to reactor closures orreactor coolant shutoff valves, steam shutoff valves, etc.; a change inone or more breaker positions (e.g., reactor coolant pump power supplybreakers, steam turbine-generator output breakers, etc.); or the like.Other control actions will be apparent to persons skilled in the artbased on the teachings herein. The determined control action may bedisplayed to a user.

At block 2140, a control action for the nuclear reactor is performed. Asdescribed above, this performance may be automatic or manual.

The method stops at block 2140.

Referring now to FIG. 22, an illustrative method 2200 is provided forsimulating and/or controlling a nuclear reactor. In an embodiment,method 2200 is used to perform block 2105 above. The method 2200 startsat a block 2205.

At block 2205, a plurality of potential neighbors is identified.Continuing the description of block 2205, potential neighbors may belimited to materials that are fission products of certain isotopes,perhaps induced by incident particles of a certain energy. In anembodiment, potential neighbors for a given principal may be limited tofission products under the same “hump” of a fission yield curve orcurves as the principal. The fission yield curve of interest might be,for example, the curve of one particular fissile material's fissionreaction or the curves of multiple fission reactions in any combinationof incident particle energy and fissile material. Potential neighborsmay also be limited to materials which are characterized to the extentnecessary to be suitable as agent materials. In an embodiment, theplurality of potential neighbors for a given principal is chosen byidentifying some number (e.g., three) of agent materials having atomicmass numbers (A) “most similar” to that of the principal material. The“most similar” decision may be restricted to agent materials havinglarger (or smaller) atomic mass numbers. Also, the “most similar”decision may be forced to take at least one smaller and one larger (inatomic mass number) agent material. Potential neighbors having amicroscopic cross section of zero or close to zero may be ruled out insome embodiments.

At block 2210, a neighbor is selected from the plurality of potentialneighbors. Once a plurality of potential neighbors is identified, one ormore neighbors may be selected from the plurality. In an embodiment,neighbors may be selected by comparing one or more microscopicproperties such as a cross section. Thus, the number of comparisonsneeded to select a neighbor is limited by the number of materialsdetermined to be potential neighbors in block 2205.

The method stops at block 2210. In an embodiment, potential neighborsmay be identified by comparing one or more microscopic properties suchas a cross section. Potential neighbors having a microscopic crosssection of zero or close to zero may be ruled out in some embodiments.

Enhanced Neutronics Modeling

Now that illustrative embodiments of nuclear reactors and reactorcontrol and simulation have been discussed, including movement andmapping of materials in nuclear reactors, illustrative systems andmethods associated with enhanced neutronics modeling will now bediscussed.

There are a wide variety of conventional codes in use for simulation andmodeling of nuclear reactor performance. Fast reactor cross sectionprocessing codes include, for example, ETOE-2, MC2-2, SDX. Diffusion andtransport theory codes include, for example, DIF3D, DIF3DK, VARIANT, andVIM. Fuel cycle/depletion codes include, for example, REBUS-3, RCT andORIGEN-RA. Perturbation theory codes include, for example, VARI3D.Thermal-hydraulic codes include, for example, SE2-ANL (SUPERENERGY2).Reactor dynamics and safety analysis codes include, for example, SAS4A,SASSYS-1 and SAS-DIF3DK. Surveillance and diagnostics codes include, forexample, MSET and PRODIAG. Stochastic, Monte-Carlo based neutronicsmodeling codes include, for example, KENO, MONK and various versions ofMCNP.

The enhanced neutronics modeling in the embodiments described belowallows for the creation, maintenance and storage of a standardized setof data describing the state of a nuclear reactor under input conditionsestablished by an operator. The state of the reactor is stored as anabstract nuclear reactor model (ANRM). The abstract nuclear reactormodel can be created and maintained regardless of the particular mode ofsimulation employed by the enhanced neutronics modeling scheme.Maintenance of an abstract nuclear reactor model allows for a number ofimprovements over conventional, and typically proprietary, neutronicsmodeling schemes.

For example, standardized data reflecting the state of the nuclearreactor allows the information to be easily accessed in formatunderstandable to programmers, modelers, and reactor operators.Importantly, maintenance of an abstract nuclear reactor modelcharacterized with standardized data sets also allows the ability totransform the data into a variety of data structures useable in other,different neutronics modeling schemes for verification. Enabling simpleverification of neutronics simulation results across multiple modelingprograms greatly improves reliability—an important feature, forinstance, when attempting to secure funding for a multi-billion dollarnuclear reactor project. Conventional analysis typically does not allowthe raw input data from one particular modeling program to be readilyused in another. Such verification analysis has previously been hamperedby a lack of a standardized data format.

As illustrated, a wide variety of codes currently exist for nuclearreactor simulation and modeling. It is also not uncommon for known codeto form the basis for multiple, subsequent, proprietary versions. Theresult is a plethora of nuclear reactor simulation and modelingprograms. They are often proprietary, and they have little or no knowninteroperability or standardization. Further, the front end ofconventional modeling programs (i.e., the means by which the simulationinput conditions are described and generated) are typically cumbersomeand are within the capability of only the most experienced users. Stillfurther, verification of nuclear reactor simulations, which requiressubstantially similar results using different programs, is expensive,time consuming, and (at worst) unreliable. This may be unacceptable whenbillion-dollar investments hang in the balance. There is a need,therefore, for a modeling interface capable of creating standardizeddata sets that can be used in creating and maintaining an abstractnuclear reactor model. One novel simulation and modeling interface willbe described supporting deterministic-type modeling.

A Deterministic Modeling Interface

One exemplary enhanced neutronics modeling scheme employs a modelinginterface for deterministic neutronics modeling. As noted above, anexemplary deterministic modeling program is REBUS-3. REBUS-3 is a systemof codes designed for the analysis of reactor fuel cycles. Two basictypes of problems are typically solved by REBUS-3: 1) the infinite-time,or equilibrium, conditions of a nuclear reactor operating under a fixedfuel management scheme; or, 2) the explicit cycle-by-cycle, ornonequilibrium operation of a reactor under a specified periodic ornon-periodic fuel management program. For the equilibrium type problems,the code uses specified external fuel supplies to load the reactor.Optionally, reprocessing may be included in the specification of theexternal fuel cycle and discharged fuel may be recycled back into thereactor. For non-equilibrium cases, the initial composition of thereactor core may be explicitly specified or the core may be loaded fromexternal feeds and discharged fuel may be recycled back into the reactoras in equilibrium problems.

A novel modeling interface is described that analyzes received reactormodeling data and nuclear reactor simulation data to create and maintainan abstract nuclear reactor model (ANRM).

By way of illustration, FIG. 23 shows a nuclear reactor modeling system2300 comprising of a modeling interface 2310, nuclear reactor modelingdata 2320, simulation data 2340, and a database 2360. The modeling data2320 further includes a plurality of data types 2330-1 to 2330-n. Forexample, the modeling data 2320 may include nuclear reactor materialdata 2330-1, geometry data 2330-2 of a portion of the nuclear reactormodel or nuclear reactor performance data 2330-n or some portionthereof.

In an embodiment, the nuclear reactor material data 2330-1 may includefuel data, structural data, shielding data, coolant data, isotope data,moderator data, and cycle load data. In an embodiment, the nuclearreactor performance data 2330-n may relate to fuel cell swelling, fueldepletion, fission product removal, coolant expulsion and fission gasremoval for all, or a portion, of the nuclear reactor under simulation.

The simulation data 2340 is generated by one or more simulators 2350-1to 2350-n. For example, the simulators could include a neutronicssimulator 2350-1, a fuel burn simulator 2350-2, or a thermal hydraulicsimulator 2350-n. The simulation data 2340 may also be generated by amaterial performance simulator, a thermal simulator or an atomisticsimulator. In an embodiment, the neutronics simulator 2350-1 is astochastic simulation tool. In a further embodiment, the stochasticsimulation tool is based on a Monte Carlo N-Particle transport code(MCNP) simulation tool. In an embodiment, the neutronics simulator2350-1 may also be a deterministic simulation tool. For example, thedeterministic simulation tool is a REBUS simulation tool. In anembodiment, the neutronics simulator 2350-1 interacts with the fuel burnsimulator 2350-2 to iteratively produce time dependent nuclear reactorsimulation data.

The modeling interface 2310 receives the modeling data 2320 as input,sends the modeling data 2320 to any number of simulators 2350-1 to2350-n, and receives the output simulation data 2340. In an embodiment,the modeling interface 2310 builds an abstract nuclear reactor model(ANRM) 2362 from analyzing both the modeling data 2320 and thesimulation data 2340. In an embodiment, the simulation data 2340 mayinclude embedded metadata added by the simulators 2350-1 to 2350-n todetermine an additional state of the ANRM 2362. The ANRM 2362 is made upof homogenized assemblies which are made up of axial blocks as will bediscussed in further detail. The ANRM 2362 is stored in database 2360for either subsequent analysis or viewing by the user. In an embodiment,the modeling interface 2310 standardizes the data representing the ANRM2362 defining structural, behavioral or creational patterns in an objectorientated program environment that are sufficient to describe a certainstate of the ANRM 2362 as will be discussed in further detail. In anembodiment, once the ANRM 2362 has been created, the modeling interface2310 may run subsequent cycles, perform safety coefficient generation,run other coupled physics codes (such as thermal hydraulics or fuelperformance) and/or produce succinct summaries.

As previously mentioned, the modeling interface 2310 uses an objectorientated programming environment to build the ANRM 2362 based oninputted data. In an example, object orientated programs include datastructures or classes wherein each instance of the class is an object.Various functions can be called upon to retrieve, modify, or add data toeach object. Examples of commonly used object orientated programsinclude C++, Python, and Java.

By way of illustration, FIG. 24A shows the fundamental class structure2400 of the modeling interface 2410. The class structure 2400 comprises:a main operator 2410, a nuclear reactor data structure 2420, an assemblylevel 2430 comprising of individual assembly structures 2440, and ablock level 2450 comprising of individual block structures 2460.Referring to FIG. 24B, in an embodiment, each block structure 2460 isgeometrically arranged within an assembly structure 2440. In anembodiment, each block structure 2460 includes one or more materialvariables. For example, the material variables can include density,flux, power, temperature, and flow. Referring to FIG. 24C, in anembodiment, a single block structure 2460 may include a plurality oflocations 2470 for which material variables are stored.

The main operator 2410 reads all inputs and builds the ANRM 2362 whichis the nuclear reactor data structure 2420 in a certain state. In anembodiment, the main operator 2410 may modify the input by addingmetadata for use in determining a first or additional state of the ANRM2362. In an embodiment, the main operator 2410 further controls andprocesses typical multi-cycle coupled simulations, fuel performancelookups, history tracking, summary making, fuel management, databaseinteraction and restarts.

The nuclear reactor data structure 2420 contains the state of thereactor at any given time. The nuclear reactor data structure 2420comprises, for example, one or more assembly structures 2440. Dataregarding the state of the reactor at any time can be stored in thedatabase 2360. In an embodiment, the nuclear reactor data structure 2420could further comprise a spent fuel pool structure (not shown). In anembodiment, the nuclear reactor data structure 2420 could furthercomprise a fuel handling machine structure (not shown).

An assembly structure 2440 comprises one or more block structures 2460.The history of an assembly structure 2440 is produced at the end of thebuilding of the ANRM 2362. The history summary may be Lagrangian innature, following a specific assembly through its path. In anembodiment, an assembly structure 2440 can act like a list and beiterated over or indexed.

A block structure 2460 contains the majority of the simulation data andmaterial variables. The histories of both the block structure 2460 andeach location 2470 within the block structure 2460 are produced at theend of the building of the ANRM 2362. The history summary of the blockstructure 2460 may also be Lagrangian in nature while the historysummary of the location 2470 within the block structure 2460 may beEulerian, which is focused on a specific spatial location as assembliespass through it.

Referring now to FIG. 25, an example of a file 2500 to be received bythe modeling interface 2310 as input modeling data 2320 is displayed. Inan embodiment, the file contains geometry descriptions and locations ofeach assembly and specifies composition labels of the assemblies thatcorrespond to nuclide-level loading labels. For example, file 2500 showsa reactor 2510 that includes three exemplary assemblies, P0001, E0002,and E0003. Each assembly is shown to further include exemplarycomposition and geometry data for each block within the assembly. In anembodiment, the file is written in an XML format. Other textual dataformats may be used to input modeling data, for example, XHTML, RSS,Atom, and KML.

Referring now to FIG. 26, an example of an input graphical userinterface (GUI) 2600 used to input modeling data 2320 to the modelinginterface 2310 is displayed. In an embodiment, the input GUI 2600further allows the user to choose which simulators to use and theparameters for each simulation. In an embodiment, the input GUI 2600includes a plurality of tabs which allow the user to input variousparameters. For example, the tabs may include a simulation parameter tab2610, a reactor parameter tab 2620, a safety calculation tab 2630, aREBUS settings tab 2640, an MCNP tab 2650, a fuel management tab 2670,and a tab for other settings 2660. In an example, the fuel managementtab 2670 may include parameter fields regarding how to move and organizethe fuel rods around at each cycle. A new job designed to create theANRM 2362 may be executed from the exemplified input GUI 2600.

The code that makes up the entirety of the modeling interface 2310 canbe divided into subgroups of functions. For example, the subgroups mayinclude components, modules, and external code interfaces. Table 1 belowdisplays a listing of at least some of the exemplary components includedin the code of the modeling interface 2310. In an embodiment, thecomponents contain the basic ANRM 2362 which include the blockstructures 2660 and the assembly structures 2640. The bulk of themodeling interface code is included in the components.

TABLE 1 Exemplary listing of modeling interface components NameDescription assemblies.py Contains Assembly class and many subclassesthat deal with Assembly level methods. assemblyLists.py ContainsAssemblyList classes that are useful for tracking assemblies that changeor discharge from the core. blocks.py Contains Block class and all itsmethods. database.py Contains Database and Database borg objects thatinterface with the SQL database for data persistence. fuelHandlers.pyContains FuelHandler object and code relevant to moving fuel around thereactor. operatorFactory.py Contains compositions of the overall mainclasses. This is where to import from to run a full modeling interfacecase. historyTracker.py Contains HistoryInterface that tracks block andassembly parameters for requested blocks and assemblies. Producesreports of these objects as they move through the core. interfaces.pyContains the base class for all Interfaces. library.py Contains codethat can read ISOTXS binary libraries and other such things like SPECTR,etc. locations.py Location object code and subclasses. Location objectstell assemblies and blocks where they are in terms of rows/positions,coordinates, indices, etc. Can also find symmetric identicals.armiLog.py Performs logging operation to print to the standard output.nucDirectory.py Contains nuclide data like atomic mass, density, Znumber, etc. operators.py Contains the main loop and main operators thatcoordinate most of the modeling interface code. paramSearch.py Containscode that allows many similar cases with a few parameters to besubmitted sequentially for parameter search studies. reactors.pyContains the Reactor class, defining the main object that holds theassembly objects settings.py Contains the case settings object thattravels around with all the global settings like power level, etc.skeletalInputs.py Contains large amounts of input that are copieddirectly or slightly modified into MC**2/REBUS inputs such as burnchains and fission product yields. sodiumRemoval.py Controls sodium bondremoval fuel performance coupling. submitter.py The GUI modelinginterface Submission Control Program. Click this to start a modelinginterface run. summarizer.py An interface that produces useful summariesduring a run. twr_shuffle.py Handles input parameters and instantiatesthe proper operator. Utils.py Contains some basic utilities like filecleaning and list searching.

Table 2 below displays a table with a listing of at least some of theexemplary modules included in the modeling interface code. In anembodiment, the modules contain code that adds physics to the modelinginterface 2310 using internal codes. In an example, the thermo moduleadds temperatures and flow rates to a model that already has power andflux. In another example, the safety coefficient generator module runsthe modeling interface in a manner that produces reactivitycoefficients.

TABLE 2 Exemplary listing of modeling interface modules Name DescriptionerWorth.py Contains code with specialized operator that inserts controlrods and produces control rod worth curves. reprocessing.py Containscode that can modify assemblies as a reprocessing plant would, flippingthem, refining them, etc. safetyCoefficients.py The safety coefficientgenerator. waveBuilder.py Contains logic that, given an equilibriumcase, will try to build enrichment distribution that matches it,building the wave in place, if you will. thermo.py Thermal-hydraulicmodel that adds flow orificing and temperature distributions to themodeling interface state after power has been computed.

Table 3 below displays a listing of at least some of the exemplaryexternal code interfaces included in the modeling interface code. In anembodiment, the external code interfaces are the links between themodeling interface 2310 and the simulators 2350-1 to 2350-n. Examples ofsimulator programs include, but are not limited to, REBUS, SASSYS,SUPERENERGY, and FEAST. In an embodiment, the external code interfacesmay be called upon by the main operator 2610.

TABLE 3 Exemplary listing of modeling interface external code interfacesName Description eqRebusInterface.py Interfaces with the REBUS usingfast equilibrium search capabilities built into REBUS. Can getequilibrium state much faster than running many explicit cycles.feastExtract.py Extracts FEAST relevant fuel performance informationfrom MCNPXT r files. feastInterface.py Interacts with modeling interfaceto extract fluxes, powers, etc, and builds FEAST input files forselected assembles. Can also run FEAST. mcnpInterface.py Allows modelinginterface to create MCNP/MCNPXT inputs with shuffling. This containsability to make homogenized hexes/triangles and pin-detail models.povRayInterface.py Interacts with POVRay ray tracer to producehigh-quality 3-D plots of the core. rebusInterface.py Containssubclasses of Operator and Reactor, etc. that has specific functionalitytied to REBUS and MC**2. This produces the inputs, runs the codes, andcalls the output-reading classes. rebusOutputs.py Contains the classesthat can read and abstract REBUS related output files.sassysInterface.py Contains an interface and Input Writers/OutputReaders for SASSYS transient analysis code. superEnergyInterface.pyInteracts with SuperEnergy to produce more detailed thermal hydraulicinformation.

Referring now to FIG. 27, an example of an output GUI 2700 is displayedfor multi-dimensional visualization of the current state of the ANRM2362 stored in the database 2360. In an embodiment, the exemplary GUI2700 can be used to read the material variables from any block structure2660 stored in the database 2360 in an explorable manner. In anembodiment, the exemplary GUI can be used to interact in real time withthe abstract nuclear reactor model by either modifying/receiving themodeling data 2320 or directing the analysis of the modeling data 2320and the simulation data 2340. For example, a block level view 2710 isdisplayed whereby a user may obtain data from each location 2720 withinthe block structure 2660. It can be appreciated that more than one levelwithin the ANRM 2362 can be viewed simultaneously as exemplified by themultiple view windows depicted in FIG. 27.

Referring now to FIG. 28, an illustrative method 2800 is provided formaintaining and standardizing an abstract data model. The method 2800starts at block 2810.

At block 2810, modeling data which represents a first part of a systemis received. For example, the modeling data may include nuclear reactormaterial data, geometry data of a portion of the nuclear reactor modelor nuclear reactor performance data or some portion thereof, as fullydescribed above. In an example, the modeling data may be received by afile. The exemplary file can contain geometry descriptions and locationsof each assembly and specify composition labels of the assemblies thatcorrespond to nuclide-level loading labels, as fully described above. Inan embodiment, each assembly can be shown to further include exemplarycomposition and geometry data for each block within the assembly.

In embodiments, the file may written in an XML format. As previouslynoted, other textual data formats may be used to input modeling data,for example, XHTML, RSS, Atom, and KML. In another example the modelingdata can be received through a GUI. In an embodiment, the GUI caninclude a plurality of tabs which allow the user to input variousparameters.

At block 2820, simulation data from a simulator capable of simulating afirst part of the system from a first set of simulation parameters isreceived. As described above, examples of simulators can include aneutronics simulator, a fuel burn simulator, a thermal hydraulicsimulator, a material performance simulator, a thermal simulator or anatomistic simulator.

At block 2830, the modeling data and the simulation data are analyzed.In an embodiment, the analysis is performed by the modeling interface.At block 2840, intermediate data is generated representing a second partof the system. In an embodiment, the intermediate data is generated bythe modeling interface. In an embodiment, the intermediate datarepresents a state of a nuclear reactor determined by the analysis ofthe modeling data and simulation data collected by the modelinginterface.

At block 2850, an abstract data model representing a certain systemstate and characterized by the modeling data, simulation data, andintermediate data is maintained. In an embodiment, the abstract datamodel is stored in a nuclear reactor data structure, as previouslydiscussed, further comprising of assembly structures and blockstructures.

At block 2860, the data representing the system state is standardized.In an embodiment, the standardized data contains defined structural,behavioral or creational patterns in an object orientated programenvironment that are sufficient to describe a certain state of theabstract data model. At block 2870, the data is exported to a database.In an embodiment, the data can be maintained for subsequent systemanalysis once it has been exported to the database. In an example, thesubsequent system analysis may include running subsequent cycles,performing safety coefficient generation, running other coupled physicscodes (such as thermal hydraulics or fuel performance) and/or producingsuccinct summaries. In another embodiment, the data may be viewed by theuser as previously discussed.

Referring now to FIG. 29, an illustrative method 2900 is provided forcontrolling a nuclear fission reactor. It will be appreciated that thenuclear fission reactor controlled by the method 2900 may be any nuclearfission reactor as desired, such as without limitation any of theillustrative nuclear fission reactors described above.

Referring additionally to FIG. 10, it will also be appreciated that themethod 2900 suitably may implemented as computer-readable code executedon a suitable computer, such as the computer system 1000. In thisembodiment, the computer system 1000 is coupled to the Reactor ControlSystem 1030. As discussed above, the Reactor Control System 1030 may bedirectly interfaced to the communications infrastructure 1006 as shownin the figure, or the Reactor Control System 1030 may also be interfacedvia communications interface 1024 or communications interface 1024 andcommunications path 1026, as desired for a particular application.Details of the Reactor Control System 1030 are discussed above and neednot be repeated for an understanding.

The method 2900 starts at block 2910. At block 2910, modeling data whichrepresents a first part of a system is received. For example, themodeling data may include nuclear reactor material data, geometry dataof a portion of the nuclear reactor model or nuclear reactor performancedata or some portion thereof, as fully described above. In an example,the modeling data may be received by a file. The exemplary file cancontain geometry descriptions and locations of each assembly and specifycomposition labels of the assemblies that correspond to nuclide-levelloading labels, as fully described above. In an embodiment, eachassembly can be shown to further include exemplary composition andgeometry data for each block within the assembly.

In embodiments, the file may written in an XML format. As previouslynoted, other textual data formats may be used to input modeling data,for example, XHTML, RSS, Atom, and KML. In another example the modelingdata can be received through a GUI. In an embodiment, the GUI caninclude a plurality of tabs which allow the user to input variousparameters.

At block 2920, simulation data from a simulator capable of simulating afirst part of the system from a first set of simulation parameters isreceived. As described above, examples of simulators can include aneutronics simulator, a fuel burn simulator, a thermal hydraulicsimulator, a material performance simulator, a thermal simulator or anatomistic simulator.

At block 2930, the modeling data and the simulation data are analyzed.In an embodiment, the analysis is performed by the modeling interface.At block 2940, intermediate data is generated representing a second partof the system. In an embodiment, the intermediate data is generated bythe modeling interface. In an embodiment, the intermediate datarepresents a state of a nuclear reactor determined by the analysis ofthe modeling data and simulation data collected by the modelinginterface.

At block 2950, an abstract data model representing a certain systemstate and characterized by the modeling data, simulation data, andintermediate data is maintained. In an embodiment, the abstract datamodel is stored in a nuclear reactor data structure, as previouslydiscussed, further comprising of assembly structures and blockstructures.

At block 2960, the data representing the system state is standardized.In an embodiment, the standardized data contains defined structural,behavioral or creational patterns in an object orientated programenvironment that are sufficient to describe a certain state of theabstract data model. At block 2970, the data is exported to a database.In an embodiment, the data can be maintained for subsequent systemanalysis once it has been exported to the database. In an example, thesubsequent system analysis may include running subsequent cycles,performing safety coefficient generation, running other coupled physicscodes (such as thermal hydraulics or fuel performance) and/or producingsuccinct summaries. In another embodiment, the data may be viewed by theuser as previously discussed.

At a block 2980, the data is provided to a reactor control system, suchas the Reactor Control System 1030. For example, as described above, inan embodiment, the Reactor Control System 1030 may be directlyinterfaced to the computer system 1000 via the communicationsinfrastructure 1006 as shown in FIG. 10, or the Reactor Control System1030 may also be interfaced via communications interface 1024 orcommunications interface 1024 and communications path 1026, as desiredfor a particular application. It will be appreciated that, regardless ofhow the interface between the Reactor Control System 1030 and thecomputer system 1000 is implemented, the result is that the standardizeddata is provided from the database 2360 (FIG. 23) to the Reactor ControlSystem 1030.

The Reactor Control System 1030 may use the provided data to controlparameters of the controlled nuclear fission reactor (not shown) asdesired, dependent upon the system state of the reactor described in thestandardized data and the actual state of the controlled nuclear fissionreactor (not shown).

For example, in an embodiment the Reactor Control System 1030 may beimplemented as the control system 720 (FIG. 7A) and may determineappropriate corrections (positive or negative) to a local neutronicreactivity of the controlled nuclear fission reactor (e.g., to returnthe controlled nuclear fission reactor to a desired operating parameter,such as desired local temperatures during reactor operations at power)in response to the provided data. To that end, the Reactor ControlSystem 1030 may generate a control signal (e.g., control signal 724)indicative of a desired correction to local neutronic reactivity.

In another embodiment, the Reactor Control System 1030 may also controlother neutron affecting or absorbing features, such as control rodsand/or safety rods, to control and/or shut down the controlled nuclearfission reactor as desired, in response to the provided data.

In another embodiment, in response to the provided data the ReactorControl System 1030 may also generate control signals to order changesin various flows, e.g., the flow of heat absorbing material (e.g.,coolant) through the reactor or portions of the controlled nuclearfission reactor by ordering changes in reactor coolant pump operationand/or various valve positions in the reactor system, including but notlimited to reactor closures or reactor coolant shutoff valves, steamshutoff valves, etc. In an embodiment, the Reactor Control System 1030may also order changes in breaker positions (e.g., reactor coolant pumppower supply breakers, steam turbine-generator output breakers, etc.).

In some embodiments, the Reactor Control System 1030 may havetemperature inputs (e.g., control system 720 receiving input fromtemperature detectors 710) in addition to neutron detectors (e.g., tosense neutron flux to determine reactor power or local reactor power ata portion of the core), and flow and position detectors (e.g.,venturi-type flow detectors, valve position indicators, breaker positionindicators). In such embodiments, in response to the provided data theReactor Control System 1030 may control the flow of heat absorbingmaterial (e.g., coolant) through the reactor and/or portions of thereactor to control overall temperatures and local temperatures inresponse to overall reactor thermal power and/or local reactor thermalpower.

In some embodiments the Reactor Control System 1030 may also provideoperator indications and accept operator inputs. In such embodiments,the Reactor Control System 1030 receives the provided data, monitorsreactor operations, may provide some automatic control features (such aschanging flow rates and moving control rods or otherwise positioningneutron affecting or absorbing materials, which are described in moredetail elsewhere herein), displays operational parameters, accepts andexecutes operator inputs for manual control actions, and/or provides toan operator information indicative of actions (either takenautomatically or recommended for manual execution by the operator) basedupon operational data from the controlled nuclear fission reactor andthe data received from the computer system 1000.

CONCLUSION

With respect to the use of substantially any plural and/or singularterms herein, those having skill in the art can translate from theplural to the singular and/or from the singular to the plural as isappropriate to the context and/or application. The varioussingular/plural permutations are not expressly set forth herein for sakeof clarity.

While particular aspects of the present subject matter described hereinhave been shown and described, it will be apparent to those skilled inthe art that, based upon the teachings herein, changes and modificationsmay be made without departing from the subject matter described hereinand its broader aspects and, therefore, the appended claims are toencompass within their scope all such changes and modifications as arewithin the true spirit and scope of the subject matter described herein.Furthermore, it is to be understood that the claimed subject matter isdefined by the appended claims. It will be understood by those withinthe art that, in general, terms used herein, and especially in theappended claims (e.g., bodies of the appended claims) are generallyintended as “open” terms (e.g., the term “including” should beinterpreted as “including but not limited to,” the term “having” shouldbe interpreted as “having at least,” the term “includes” should beinterpreted as “includes but is not limited to,” etc.). It will befurther understood by those within the art that if a specific number ofan introduced claim recitation is intended, such an intent will beexplicitly recited in the claim, and in the absence of such recitationno such intent is present. For example, as an aid to understanding, thefollowing appended claims may contain usage of the introductory phrases“at least one” and “one or more” to introduce claim recitations.However, the use of such phrases should not be construed to imply thatthe introduction of a claim recitation by the indefinite articles “a” or“an” limits any particular claim containing such introduced claimrecitation to claims containing only one such recitation, even when thesame claim includes the introductory phrases “one or more” or “at leastone” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an”should typically be interpreted to mean “at least one” or “one ormore”); the same holds true for the use of definite articles used tointroduce claim recitations. In addition, even if a specific number ofan introduced claim recitation is explicitly recited, those skilled inthe art will recognize that such recitation should typically beinterpreted to mean at least the recited number (e.g., the barerecitation of “two recitations,” without other modifiers, typicallymeans at least two recitations, or two or more recitations).Furthermore, in those instances where a convention analogous to “atleast one of A, B, and C, etc.” is used, in general such a constructionis intended in the sense one having skill in the art would understandthe convention (e.g., “a system having at least one of A, B, and C”would include but not be limited to systems that have A alone, B alone,C alone, A and B together, A and C together. B and C together, and/or A.B, and C together, etc.). In those instances where a conventionanalogous to “at least one of A, B, or C, etc.” is used, in general sucha construction is intended in the sense one having skill in the artwould understand the convention (e.g., “a system having at least one ofA, B, or C” would include but not be limited to systems that have Aalone, B alone, C alone, A and B together, A and C together, B and Ctogether, and/or A, B, and C together, etc.). It will be furtherunderstood by those within the art that virtually any disjunctive wordand/or phrase presenting two or more alternative terms, whether in thedescription, claims, or drawings, should be understood to contemplatethe possibilities of including one of the terms, either of the terms, orboth terms. For example, the phrase “A or B” will be understood toinclude the possibilities of “A” or “B” or “A and B.” With respect tothe appended claims, those skilled in the art will appreciate thatrecited operations therein may generally be performed in any order.Examples of such alternate orderings may include overlapping,interleaved, interrupted, reordered, incremental, preparatory,supplemental, simultaneous, reverse, or other variant orderings, unlesscontext dictates otherwise. With respect to context, even terms like“responsive to.” “related to,” or other past-tense adjectives aregenerally not intended to exclude such variants, unless context dictatesotherwise.

The herein described subject matter sometimes illustrates differentcomponents contained within, or connected with, different othercomponents. It is to be understood that such depicted architectures aremerely exemplary, and that in fact many other architectures may beimplemented which achieve the same functionality. In a conceptual sense,any arrangement of components to achieve the same functionality iseffectively “associated” such that the desired functionality isachieved. Hence, any two components herein combined to achieve aparticular functionality can be seen as “associated with” each othersuch that the desired functionality is achieved, irrespective ofarchitectures or intermedial components. Likewise, any two components soassociated can also be viewed as being “operably connected”,“operatively coupled,” or “operably coupled.” to each other to achievethe desired functionality, and any two components capable of being soassociated can also be viewed as being “operably couplable,” to eachother to achieve the desired functionality. Specific examples ofoperably couplable include but are not limited to physically mateableand/or physically interacting components, and/or wirelesslyinteractable, and/or wirelessly interacting components, and/or logicallyinteracting, and/or logically interactable components.

While various aspects and embodiments have been disclosed herein, otheraspects and embodiments will be apparent to those skilled in the art.The various aspects and embodiments disclosed herein are for purposes ofillustration and are not intended to be limiting, with the true scopeand spirit being indicated by the following claims.

1.-208. (canceled)
 209. A computer system, comprising: at least one processor configured to perform acts of: generating a user interface in response to execution of instructions, the user interface comprising: identifying an input definition of a reactor model including a plurality of defined assemblies represented as a hierarchical data structure; within the user interface, presenting a plurality of simulator modules and permitting a user to selectively choose one or more of the plurality of simulator modules to be simulated; simulating the defined reactor model associated with the input definition including the chosen one or more of the plurality of simulator modules, generating simulation data, wherein the act of simulating the defined reactor model includes calling interfaces of the chosen one or more of the plurality of simulator modules in a predetermined sequence; and presenting an output of the act of simulating the defined reactor model within the user interface, including the generated simulation data.
 210. The computer system according to claim 209, wherein the act of presenting an output comprises an act of displaying a representation of the simulation data within the modeling interface.
 211. The computer system according to claim 210, wherein the act of presenting an output comprises an act of displaying a state of the defined reactor model within the modeling interface.
 213. The computer system according to claim 209, wherein the act of generating a user interface further comprises an act of presenting, within the user interface, a control permitting the user to provide cycle information defining at least a number and length of a cycle to be simulated by the act of simulating.
 214. The computer system according to claim 209, wherein the plurality of simulator modules comprises a neutronics simulator module, a fuel burn simulator module, a thermal hydraulics simulator module, and a material performance simulator module.
 215. The computer system according to claim 209, further comprising an act of presenting, within the user interface, a control permitting the user to provide fuel movement information and modifying the reactor model responsive to the provided information.
 216. The computer system according to claim 209, wherein the act of generating a user interface further comprises an act of providing, within the user interface, a control permitting the user to specify a target k-eff of the simulation.
 217. The computer system according to claim 209, further comprising an act of storing a data structure defining the reactor model including the plurality of defined assemblies.
 218. The computer system according to claim 217, further comprising an act of representing, for each of the defined assemblies within the reactor model, a geometry description and location.
 219. The computer system according to claim 209, wherein the hierarchical data structure includes a nuclear reactor data structure representing a nuclear reactor core, the nuclear reactor data structure including a plurality of assembly structures, each assembly structure representative of a physical component that is present in the nuclear reactor model, two or more of the assembly data structures representing fuel assemblies of the nuclear reactor core, each fuel assembly data structure including a plurality of block data structures representative of axially distributed blocks of the fuel assembly of the nuclear reactor core, at least one block structure of the plurality of block structures of the fuel assembly structure includes material data and location, the material and location data representative of fuel pin material and location within the fuel assembly of the nuclear reactor core.
 220. The computer system according to claim 219, wherein the at least one processor is adapted to convert the input definition of a reactor model to nuclear reactor modeling data defining a nuclear reactor model modelling the nuclear reactor core, the nuclear reactor modeling data including: a plurality of cell data for a plurality of cells of the nuclear reactor model, each cell defined by one or both of bounding of surfaces and regions of space, each cell data including a physical location, a material identifier, and a geometry of the associated cell of the nuclear reactor model; nuclear reactor performance data including fuel cell swelling and fuel depletion; and nuclear material data including cycle load data.
 221. The system according to claim 209, wherein the act of generating a user interface further comprises an act of presenting, within the user interface, a visual representation of the reactor model including the plurality of defined assemblies.
 222. The system according to claim 221, wherein the act of presenting, within the user interface, the visual representation of the reactor model including the plurality of defined assemblies further comprises and act of displaying the plurality of defined assemblies as a collection of components within the user interface.
 223. The system according to claim 222, wherein the collection of components are adapted indicate, to the user within the user interface, simulation data relating to each one of the plurality of defined assemblies.
 224. The system according to claim 222, wherein the collection of components within the user interface are selectable by a user to view information relating to a series of block structures relating to a particular one of the plurality of defined assemblies.
 225. The system according to claim 224, wherein the series of block structures each include information identifying a state of material within the particular one of the plurality of defined assemblies. 